Sample records for zz-fsxlib-jd99 mcnp nuclear

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tallymore » characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data.« less

Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suitesmore » were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.« less

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capabilitymore » of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.« less

During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.

Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

We measured the skyshine dose from turbine buildings at Shimane Nuclear Power Station Unit 1 (NS-1) and Unit 2 (NS-2), and then compared it with the dose calculated with the Monte Carlo transport code MCNP5. The skyshine dose values calculated with the MCNP5 code agreed with the experimental data within a factor of 2.8, when the roof of the turbine building was precisely modeled. We concluded that our MCNP5 calculation was valid for BWR turbine skyshine dose evaluation.

We, the development teams for MCNP, NJOY, and parts of ENDF, would like to invite you to a proposed 3 day workshop October 30, 31 and November 1 2012, to be held at Los Alamos National Laboratory. At this workshop, we will review new and developing missions that MCNP6 and the underlying nuclear data are being asked to address. LANL will also present its internal plans to address these missions and recent advances in these three capabilities and we will be interested to hear your input on these topics. Additionally we are interested in hearing from you additional technical advances,more » missions, concerns, and other issues that we should be considering for both short term (1-3 years) and long term (4-6 years)? What are the additional existing capabilities and methods that we should be investigating? The goal of the workshop is to refine priorities for mcnp6 transport methods, algorithms, physics, data and processing as they relate to the intersection of MCNP, NJOY and ENDF.« less

MCNPX simulations have been used to guide the development of a portable inspection system for narcotics, explosives, and special nuclear material (SNM) detection. The system seeks to address these threats to national security by utilizing a high-yield, compact neutron source to actively interrogate the threats and produce characteristic signatures that can then be detected by radiation detectors. The portability of the system enables rapid deployment and proximity to threats concealed in small spaces. Both dD and dT electronic neutron generators (ENG) were used to interrogate ammonium nitrate fuel oil (ANFO) and cocaine hydrochloride, and the detector response of NaI, CsI, and LaBr3 were compared. The effect of tungsten shielding on the neutron flux in the gamma ray detectors was investigated, while carbon, beryllium, and polyethylene ENG moderator materials were optimized by determining the reaction rate density in the threats. In order to benchmark the modeling results, experimental measurements are compared with MCNPX simulations. In addition, the efficiency and die-away time of a portable differential die-away analysis (DDAA) detector using 3He proportional counters for SNM detection has been determined.

This report is intended to the MCNP6 developers and sponsors of MCNP6. It presents a set of suggested possible future improvements to MCNP6 and to its CEM03.03 and LAQGSM03.03 event-generators. A few suggested modifications of MCNP6 are quite simple, aimed at avoiding possible problems with running MCNP6 on various computers, i.e., these changes are not expected to change or improve any results, but should make the use of MCNP6 easier; such changes are expected to require limited man-power resources. On the other hand, several other suggested improvements require a serious further development of nuclear reaction models, are expected to improvemore » significantly the predictive power of MCNP6 for a number of nuclear reactions; but, such developments require several years of work by real experts on nuclear reactions.« less

In 2002 and again in 2003, an investigative journalist unit at ABC News transported a 6.8 kilogram metallic slug of depleted uranium (DU) via shipping container from Istanbul, Turkey to Brooklyn, NY and from Jakarta, Indonesia to Long Beach, CA. Targeted inspection of these shipping containers by Department of Homeland Security (DHS) personnel, included the use of gamma-ray imaging, portal monitors and hand-held radiation detectors, did not uncover the hidden DU. Monte Carlo analysis of the gamma-ray intensity and spectrum of a DU slug and one consisting of highly-enriched uranium (HEU) showed that DU was a proper surrogate for testing the ability of DHS to detect the illicit transport of HEU. Our analysis using MCNP-5 illustrated the ease of fully shielding an HEU sample to avoid detection. The assembly of an Improvised Nuclear Device (IND) -- a crude atomic bomb -- from sub-critical pieces of HEU metal was then examined via Monte Carlo criticality calculations. Nuclear explosive yields of such an IND as a function of the speed of assembly of the sub-critical HEU components were derived. A comparison was made between the more rapid assembly of sub-critical pieces of HEU in the ``Little Boy'' (Hiroshima) weapon's gun barrel and gravity assembly (i.e., dropping one sub-critical piece of HEU on another from a specified height). Based on the difficulty of detection of HEU and the straightforward construction of an IND utilizing HEU, current U.S. government policy must be modified to more urgently prioritize elimination of and securing the global inventories of HEU.

For many real-world applications in radiation transport where simulations are compared to experimental measurements, like in nuclear criticality safety, the bias (simulated - experimental k eff) in the calculation is an extremely important quantity used for code validation. The objective of this project is to accurately predict the bias of MCNP6 [1] criticality calculations using machine learning (ML) algorithms, with the intention of creating a tool that can complement the current nuclear criticality safety methods. In the latest release of MCNP6, the Whisper tool is available for criticality safety analysts and includes a large catalogue of experimental benchmarks, sensitivity profiles,more » and nuclear data covariance matrices. This data, coming from 1100+ benchmark cases, is used in this study of ML algorithms for criticality safety bias predictions.« less

This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence inmore » the simulation is acquired.« less

If a severe accident with a considerable core damage occurs at a nuclear power plant whose containment top is remarkably thin compared with the walls, the radiation transported through the top and scattered in air may cause high dose rates at the power plant area. Noble gases and other fission products released to the containment act as sources. The dose rates caused by skyshine have been calculated by MCNP3A for the Loviisa nuclear power plant (two-unit, 445-MW VVER) for the outside area and inside some buildings, taking the attenuation in the roofs of the buildings into account.

To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicabilitymore » for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.« less

The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore » predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less

With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

This presentation describes how to build MCNP 6.2. MCNP®* 6.2 can be compiled on Macs, PCs, and most Linux systems. It can also be built for parallel execution using both OpenMP and Messing Passing Interface (MPI) methods. MCNP6 requires Fortran, C, and C++ compilers to build the code.

The discussion below provides a status report on the MCNP 2020 initiative. It includes discussion of the history of MCNP 2020, accomplishments during 2013-17, priorities for near-term development, other related efforts, a brief summary, and a list of references for the plans and work accomplished.

Monte Carlo N-Particle or MCNP ® is a general-purpose Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. This MCNP Version 6.2 follows the MCNP6.1.1 beta version and has been released in order to provide the radiation transport community with the latest feature developments and bug fixes for MCNP. Since the last release of MCNP major work has been conducted to improve the code base, add features, and provide tools to facilitate ease of use of MCNP version 6.2 as well as the analysis of results. These release notes serve as a general guidemore » for the new/improved physics, source, data, tallies, unstructured mesh, code enhancements and tools. For more detailed information on each of the topics, please refer to the appropriate references or the user manual which can be found at http://mcnp.lanl.gov. This release of MCNP version 6.2 contains 39 new features in addition to 172 bug fixes and code enhancements. There are still some 33 known issues the user should familiarize themselves with (see Appendix).« less

This paper describes MCNP, states its philosophy, introduces a number of new features becoming available with version MCNP4A, and answers a number of questions asked by participants in the workshop. MCNP is a general-purpose three-dimensional neutron, photon and electron transport code. Its philosophy is ``Quality, Value and New Features.`` Quality is exemplified by new software quality assurance practices and a program of benchmarking against experiments. Value includes a strong emphasis on documentation and code portability. New features are the third priority. MCNP4A is now available at Los Alamos. New features in MCNP4A include enhanced statistical analysis, distributed processor multitasking, newmore » photon libraries, ENDF/B-VI capabilities, X-Windows graphics, dynamic memory allocation, expanded criticality output, periodic boundaries, plotting of particle tracks via SABRINA, and many other improvements. 23 refs.« less

A neutron irradiator has been assembled at IPEN facilities to perform qualitative-quantitative analysis of many materials using thermal and fast neutrons outside the nuclear reactor premises. To establish the prototype specifications, the neutron flux distribution and the absorbed dose rates were calculated using the MCNP computer code. These theoretical predictions then allow one to discuss the optimum irradiator design and its performance.

The weight window variance reduction method in the general-purpose Monte Carlo N-Particle radiation transport code MCNPTM has recently been rewritten. In particular, it is now possible to generate weight window importance functions on a superimposed mesh, eliminating the need to subdivide geometries for variance reduction purposes. Our assessment addresses the following questions: (1) Does the new MCNP4C treatment utilize weight windows as well as the former MCNP4B treatment? (2) Does the new MCNP4C weight window generator generate importance functions as well as MCNP4B? (3) How do superimposed mesh weight windows compare to cell-based weight windows? (4) What are the shortcomingsmore » of the new MCNP4C weight window generator? Our assessment was carried out with five neutron and photon shielding problems chosen for their demanding variance reduction requirements. The problems were an oil well logging problem, the Oak Ridge fusion shielding benchmark problem, a photon skyshine problem, an air-over-ground problem, and a sample problem for variance reduction.« less

In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k eff.

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members acceptmore » the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk eff (ISG-8, Rev. 3, Recommendation 4).« less

One of the most difficult tasks when analyzing a complex three-dimensional system with Monte Carlo is geometry model development. SABRINA attempts to make the modeling process more user-friendly and less of an obstacle. It accepts both combinatorial solid bodies and MCNP surfaces and produces MCNP cells. The model development process in SABRINA is highly interactive and gives the user immediate feedback on errors. Users can view their geometry from arbitrary perspectives while the model is under development and interactively find and correct modeling errors. An example of a SABRINA display is shown. It represents a complex three-dimensional shape.

The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy.more » Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation. Presented in this paper is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay

In MCNP6.1, materials are constructed by identifying the constituent isotopes (or elements in a few cases) individually. This list selects the corresponding microscopic cross sections calculated from the free-gas model to create the material macroscopic cross sections. Furthermore, the free-gas model and the corresponding material macroscopic cross sections assume that the interactions of atoms do not affect the nuclear cross sections.

In order to reduce the user's time and the computer time needed to solve deep penetration problems, an automated variance reduction capability has been developed for the MCNP Monte Carlo transport code. This new variance reduction capability developed for MCNP5 employs the PARTISN multigroup discrete ordinates code to generate mesh-based weight windows. The technique of using deterministic methods to generate importance maps has been widely used to increase the efficiency of deep penetration Monte Carlo calculations. The application of this method in MCNP uses the existing mesh-based weight window feature to translate the MCNP geometry into geometry suitable for PARTISN. The adjoint flux, which is calculated with PARTISN, is used to generate mesh-based weight windows for MCNP. Additionally, the MCNP source energy spectrum can be biased based on the adjoint energy spectrum at the source location. This method can also use angle-dependent weight windows.

An MCNP 3-dimensional model can be widely used to evaluate various design parameters such as a core design or shielding design. Conventionally, a simplified 3-dimensional MCNP model is applied to calculate these parameters because of the cumbersomeness of modelling by hand. ANSYS has a function for converting the CAD `stp' format into an MCNP input in the geometry part. Using ANSYS and a 3- dimensional CAD file, a very detailed and sophisticated MCNP 3-dimensional model can be generated. The MCNP model is applied to evaluate the assembly weighting factor at the ex-core detector of SMART, and the result is compared with a simplified MCNP SMART model and assembly weighting factor calculated by DORT, which is a deterministic Sn code.

MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way. New version program summaryProgram title: MODAR Catalogue identifier: AEGA_v1_1 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_1.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 150 927 No. of bytes in distributed program, including test data, etc.: 4 981 633 Distribution format: tar.gz Programming language: C++ Computer: Most Unix workstations and PCs Operating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP. RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two dimensional 139×740 bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself. Classification: 17.6 Catalogue identifier of previous version: AEGA_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181 (2010) 1161 External routines: ROOT version 5.24.00 ( http://root.cern.ch/drupal/) Does the new version supersede the previous version?: Yes Nature of problem: The output of a MCNP simulation is an ascii file. The data processing is usually performed by copying and pasting the relevant parts of the ascii

MCNP Output Data Analysis with ROOT (MODAR) is a tool based on CERN's ROOT software. MODAR has been designed to handle time-energy data issued by MCNP simulations of neutron inspection devices using the associated particle technique. MODAR exploits ROOT's Graphical User Interface and functionalities to visualize and process MCNP simulation results in a fast and user-friendly way. MODAR allows to take into account the detection system time resolution (which is not possible with MCNP) as well as detectors energy response function and counting statistics in a straightforward way. Program summaryProgram title: MODAR Catalogue identifier: AEGA_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGA_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 155 373 No. of bytes in distributed program, including test data, etc.: 14 815 461 Distribution format: tar.gz Programming language: C++ Computer: Most Unix workstations and PC Operating system: Most Unix systems, Linux and windows, provided the ROOT package has been installed. Examples where tested under Suse Linux and Windows XP. RAM: Depends on the size of the MCNP output file. The example presented in the article, which involves three two-dimensional 139×740 bins histograms, allocates about 60 MB. These data are running under ROOT and include consumption by ROOT itself. Classification: 17.6 External routines: ROOT version 5.24.00 ( http://root.cern.ch/drupal/) Nature of problem: The output of an MCNP simulation is an ASCII file. The data processing is usually performed by copying and pasting the relevant parts of the ASCII file into Microsoft Excel. Such an approach is satisfactory when the quantity of data is small but is not efficient when the size of the simulated data is large, for example when time

Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling)more » and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.« less

Simulations of proposed material exposure experiments were performed using MCNP6. The experiments will expose ampules containing different materials of interest with radiation to observe the chemical breakdown of the materials. Simulations were performed to map out dose in materials as a function of distance from the source, dose variation between materials, dose variation due to ampule orientation, and dose variation due to different source energy. This write up is an overview of the simulations and will provide guidance on how to use the data in the spreadsheet.

This report discusses critical experiments involving SPERT-D{sup 1,2} fuel elements and BORAX-V{sup 3-8} fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

This report discusses critical experiments involving SPERT-D[sup 1,2] fuel elements and BORAX-V[sup 3-8] fuel which have been modeled and calculations performed with MCNP. MCNP is a Monte Carlo based transport code. For this study continuous-energy nuclear data from the ENDF/B-V cross section library was used. The SPERT-D experiments consisted of various arrays of fuel elements moderated and reflected with either water or a uranyl nitrate solution. Some SPERT-D experiments used cadmium as a fixed neutron poison, while others were poisoned with various concentrations of boron in the moderating/reflecting solution. ne BORAX-V experiments were arrays of either boiling fuel rod assembliesmore » or superheater assemblies, both types of arrays were moderated and reflected with water. In one boiling fuel experiment, two fuel rods were replaced with borated stainless steel poison rods.« less

The particle transport code MCNP has been used to produce a background radiation data file on a worldwide grid that can easily be sampled as a source in the code. Location-dependent cosmic showers were modeled by Monte Carlo methods to produce the resulting neutron and photon background flux at 2054 locations around Earth. An improved galactic-cosmic-ray feature was used to model the source term as well as data from multiple sources to model the transport environment through atmosphere, soil, and seawater. A new elevation scaling feature was also added to the code to increase the accuracy of the cosmic neutronmore » background for user locations with off-grid elevations. Furthermore, benchmarking has shown the neutron integral flux values to be within experimental error.« less

This study was intended to verify the capability of the Monte Carlo MCNP-4B code to evaluate spatial dose distribution based on information gathered from CT or SPECT. A new three-dimensional (3D) dose calculation approach for internal emitter use in radioimmunotherapy (RIT) was developed using the Monte Carlo MCNP-4B code as the photon and electron transport engine. It was shown that the MCNP-4B computer code can be used with voxel-based anatomic and physiologic data to provide 3D dose distributions. This study showed that the MCNP-4B code can be used to develop a treatment planning system that will provide such information in a time manner, if dose reporting is suitably optimized. If each organ is divided into small regions where the average energy deposition is calculated with a typical volume of 0.4 cm(3), regional dose distributions can be provided with reasonable central processing unit times (on the order of 12-24 h on a 200-MHz personal computer or modest workstation). Further efforts to provide semiautomated region identification (segmentation) and improvement of marrow dose calculations are needed to supply a complete system for RIT. It is envisioned that all such efforts will continue to develop and that internal dose calculations may soon be brought to a similar level of accuracy, detail, and robustness as is commonly expected in external dose treatment planning. For this study we developed a code with a user-friendly interface that works on several nuclear medicine imaging platforms and provides timely patient-specific dose information to the physician and medical physicist. Future therapy with internal emitters should use a 3D dose calculation approach, which represents a significant advance over dose information provided by the standard geometric phantoms used for more than 20 y (which permit reporting of only average organ doses for certain standardized individuals)

An exact copy of the detector model generated for Cyltran was reproduced as an MCNP input file and the detection efficiency was calculated similarly with the methodology used in previous experimental measurements and simulation of a 280 cm(3) HPGe detector. Below 1000 keV the MCNP data correlated to the Cyltran results within 0.5% while above this energy the difference between MCNP and Cyltran increased to about 6% at 4800 keV, depending on the electron cut-off energy.

An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculationsmore » were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.« less

Secondary electrons in the form of delta-rays, also referred to as knock-on electrons, have been a feature of MCNP for electron and positron transport for over 20 years. While MCNP6 now includes transport for a suite of heavy-ions and charged particles from its integration with MCNPX, the production of delta-rays was still limited to electron and positron transport. In the newest release of MCNP6, version 6.2.0, delta-ray production has now been extended for all energetic charged particles. The basis of this production is the analytical formulation from Rossi and ICRU Report 37. As a result, this paper discusses the MCNP6more » heavy charged-particle implementation and provides production results for several benchmark/test problems.« less

Secondary electrons in the form of delta-rays, also referred to as knock-on electrons, have been a feature of MCNP for electron and positron transport for over 20 years. While MCNP6 now includes transport for a suite of heavy-ions and charged particles from its integration with MCNPX, the production of delta-rays was still limited to electron and positron transport. In the newest release of MCNP6, version 6.2.0, delta-ray production has now been extended for all energetic charged particles. The basis of this production is the analytical formulation from Rossi and ICRU Report 37. This paper discusses the MCNP6 heavy charged-particle implementation and provides production results for several benchmark/test problems.

The performance of seven computer platforms has been evaluated with the MCNP4A Monte Carlo radiation transport code. For the first time we report timing results using MCNP4A and its new test set and libraries. Comparisons are made on platforms not available to us in previous MCNP timing studies. By using MCNP4A and its 325-problem test set, a widely-used and readily-available physics production code is used; the timing comparison is not limited to a single ``typical`` problem, demonstrating the problem dependence of timing results; the results are reproducible at the more than 100 installations around the world using MCNP; comparison ofmore » performance of other computer platforms to the ones tested in this study is possible because we present raw data rather than normalized results; and a measure of the increase in performance of computer hardware and software over the past two years is possible. The computer platforms reported are the Cray-YMP 8/64, IBM RS/6000-560, Sun Sparc10, Sun Sparc2, HP/9000-735, 4 processor 100 MHz Silicon Graphics ONYX, and Gateway 2000 model 4DX2-66V PC. In 1991 a timing study of MCNP4, the predecessor to MCNP4A, was conducted using ENDF/B-V cross-section libraries, which are export protected. The new study is based upon the new MCNP 25-problem test set which utilizes internationally available data. MCNP4A, its test problems and the test data library are available from the Radiation Shielding and Information Center in Oak Ridge, Tennessee, or from the NEA Data Bank in Saclay, France. Anyone with the same workstation and compiler can get the same test problem sets, the same library files, and the same MCNP4A code from RSIC or NEA and replicate our results. And, because we report raw data, comparison of the performance of other compute platforms and compilers can be made.« less

Acute radiation doses due to physical contact with a high-activity radioactive source have proven to be an occupational hazard. Multiple radiation injuries have been reported due to manipulating a radioactive source with bare hands or by placing a radioactive source inside a shirt or pants pocket. An effort to reconstruct the radiation dose must be performed to properly assess and medically manage the potential biological effects from such doses. Using the reference computational phantoms defined by the International Commission on Radiological Protection (ICRP) and the Monte Carlo N-Particle transport code (MCNP6), dose rate coefficients are calculated to assess doses for common acute doses due to beta and photon radiation sources. The research investigates doses due to having a radioactive source in either a breast pocket or pants back pocket. The dose rate coefficients are calculated for discrete energies and can be used to interpolate for any given energy of photon or beta emission. The dose rate coefficients allow for quick calculation of whole-body dose, organ dose, and/or skin dose if the source, activity, and time of exposure are known. Doses are calculated with the dose rate coefficients and compared to results from the International Atomic Energy Agency (IAEA) reports from accidents that occurred in Gilan, Iran and Yanango, Peru. Skin and organ doses calculated with the dose rate coefficients appear to agree, but there is a large discrepancy when comparing whole-body doses assessed using biodosimetry and whole-body doses assessed using the dose rate coefficients.

The objective of the research was to test available photoatomic data libraries for fusion relevant applications, comparing against experimental and computational neutronics benchmarks. Photon flux and heating was compared using the photon interaction data libraries (mcplib 04p, 05t, 84p and 12p). Suitable benchmark experiments (iron and water) were selected from the SINBAD database and analysed to compare experimental values with MCNP calculations using mcplib 04p, 84p and 12p. In both the computational and experimental comparisons, the majority of results with the 04p, 84p and 12p photon data libraries were within 1σ of the mean MCNP statistical uncertainty. Larger differences were observed when comparing computational results with the 05t test photon library. The Doppler broadening sampling bug in MCNP-5 is shown to be corrected for fusion relevant problems through use of the 84p photon data library. The recommended libraries for fusion neutronics are 84p (or 04p) with MCNP6 and 84p if using MCNP-5.

The aim of this paper is to present the capability of TRIPOLI-4®, the CEA Monte Carlo code, to model a large-scale fusion reactor with complex neutron source and geometry. In the past, numerous benchmarks were conducted for TRIPOLI-4® assessment on fusion applications. Experiments (KANT, OKTAVIAN, FNG) analysis and numerical benchmarks (between TRIPOLI-4® and MCNP5) on the HCLL DEMO2007 and ITER models were carried out successively. In this previous ITER benchmark, nevertheless, only the neutron wall loading was analyzed, its main purpose was to present MCAM (the FDS Team CAD import tool) extension for TRIPOLI-4®. Starting from this work a more extended benchmark has been performed about the estimation of neutron flux, nuclear heating in the shielding blankets and tritium production rate in the European TBMs (HCLL and HCPB) and it is presented in this paper. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model (version 4.1). Simplified TBMs (from KIT) have been integrated in the equatorial-port. Comparisons of neutron wall loading, flux, nuclear heating and tritium production rate show a good agreement between the two codes. Discrepancies are mainly included in the Monte Carlo codes statistical error.

We describe a new program to read a sequence of tomographic scans and prepare the geometry and material sections of an MCNP input file. Image processing techniques include contrast controls and mapping of grey scales to colour. The user interface provides several tools with which the user can associate a range of image intensities to an MCNP material. Materials are loaded from a library. A separate material assignment can be made to a pixel intensity or range of intensities when that intensity dominates the image boundaries; this material is assigned to all pixels with that intensity contiguous with the boundary. Material fractions are computed in a user-specified voxel grid overlaying the scans. New materials are defined by mixing the library materials using the fractions. The geometry can be written as an MCNP lattice or as individual cells. A combination algorithm can be used to join neighbouring cells with the same material.

MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo code for simulating the transport of neutrons, photons, electrons, positrons, and more recently other fundamental particles and heavy ions. Over many years MCNP has found a wide range of applications in many different fields, including medical radiation physics. In this presentation we will describe and illustrate a number of significant recently-developed features in the current version of the code, MCNP6, having particular utility for medical physics. Among these are major extensions of the ability to simulate large, complex geometries, improvement in memory requirements and speed for large lattices, introduction of mesh-basedmore » isotopic reaction tallies, advances in radiography simulation, expanded variance-reduction capabilities, especially for pulse-height tallies, and a large number of enhancements in photon/electron transport.« less

We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM)more » used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤ 7, in the case of CEM, and A ≤ 12, in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Lastly, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.« less

We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM)more » used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤ 7, in the case of CEM, and A ≤ 12, in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Lastly, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.« less

Pseudo-random number generators (PRNG) are intensively used in many stochastic algorithms in particle simulations, artificial neural networks and other scientific computation. The PRNG in Monte Carlo N-Particle Transport Code (MCNP) requires long period, high quality, flexible jump and fast enough. In this paper, we implement such a PRNG for MCNP on NVIDIA's GTX200 Graphics Processor Units (GPU) using CUDA programming model. Results shows that 3.80 to 8.10 times speedup are achieved compared with 4 to 6 cores CPUs and more than 679.18 million double precision random numbers can be generated per second on GPU.

A study was made to investigate the methods of building a simplified MCNP model for radiological dose estimation. The research was done using an example of a complicated glovebox with extra shielding. The paper presents several different calculations for neutron and photon dose evaluations where glovebox elements were consecutively excluded from the MCNP model. The analysis indicated that to obtain a fast and reasonable estimation of dose, the model should be realistic in details that are close to the tally. Other details may be omitted.

The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.

Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

performance of the detector, and to compare the performance with sodium iodide and germanium detectors. Monte Carlo ( MCNP ) simulation was used to...aluminum ~50% more efficient), and to estimate optimum shield dimensions for an HPXe based nuclear explosion monitor. MCNP modeling was also used to...detector were calculated with MCNP by using input activity levels as measured in routine NEM runs at Pacific Northwest National Laboratory (PNNL

In the framework of the REDOS project, the space-energy distribution of the neutron and photon flux has been calculated over the pressure vessel simulator thickness of the LR-0 experimental reactor, Rez, Czech Republic. The results calculated by the Monte Carlo code MCNP4C are compared with the measurements performed in the Nuclear Research Institute, Rez. The spectra have been measured at the barrel, in front of, inside and behind the pressure vessel in different configurations. The neutron measurements were performed in the energy range 0.1-10 MeV. This work has been done in the frame of the 5th Frame Work Programme of the European Community 1998-2002.

The Monte Carlo N-Particle transport code MCNP is based on a surface description of 3-dimensional geometry. Cells are defined in terms of boolean operations on signed quadratic surfaces. MCNP geometry is entered as a card image file containing coefficients of the surface equations and a list of surfaces and operators describing cells. Several programs are available to assist in creation of the geometry specification, among them Sabrina and the new ``Smart Editor`` code XSM. We briefly describe geometry creation in Sabrina and then discuss XSM in detail. XSM is under development; our discussion is based on the state of XSMmore » as of January 1, 1994.« less

The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. This paper discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models' theories are included as well.

The release of version 6.2 of the MCNP6 radiation transport code is imminent. To complement the newest release, a summary of the heavy charged particle physics models used in the 1 MeV to 1 GeV energy regime is presented. Several changes have been introduced into the charged particle physics models since the merger of the MCNP5 and MCNPX codes into MCNP6. Here, this article discusses the default models used in MCNP6 for continuous energy loss, energy straggling, and angular scattering of heavy charged particles. Explanations of the physics models’ theories are included as well.

This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

We are developing an innovative tagged neutrons inspection system (TNIS) for cargo containers: the system will allow us to assay the chemical composition of suspect objects, previously identified by a standard X-ray radiography. The operation of the system is extensively being simulated by using the MCNP Monte Carlo code to study different inspection geometries, cargo loads and hidden threat materials. Preliminary simulations evaluating the signal and the signal over background ratio expected as a function of the system parameters are presented. The results for a selection of cases are briefly discussed and demonstrate that the system can operate successfully in different filling conditions.

This report serves to extend the previous document: 'MCNP Calculations Replicating ARH-600 Data' by replicating the nitrate curves found in ARH-600. This report includes the MCNP models used, the calculated critical dimension for each analyzed parameter set, and the resulting data libraries for use with the CritView code. As with the ARH-600 data, this report is not meant to replace the analysis of the fissile systems by qualified criticality personnel. The M CNP data is presented without accounting for the statistical uncertainty (although this is typically less than 0.001) or bias and, as such, the application of a reasonable safetymore » margin is required. The data that follows pertains to the uranyl nitrate and plutonium nitrate spheres, infinite cylinders, and infinite slabs of varying isotopic composition, reflector thickness, and molarity. Each of the cases was modeled in MCNP (version 5.1.40), using the ENDF/B-VI cross section set. Given a molarity, isotopic composition, and reflector thickness, the fissile concentration and diameter (or thicknesses in the case of the slab geometries) were varied. The diameter for which k-effective equals 1.00 for a given concentration could then be calculated and graphed. These graphs are included in this report. The pages that follow describe the regions modeled, formulas for calculating the various parameters, a list of cross-sections used in the calculations, a description of the automation routine and data, and finally the data output. The data of most interest are the critical dimensions of the various systems analyzed. This is presented graphically, and in table format, in Appendix B. Appendix C provides a text listing of the same data in a format that is compatible with the CritView code. Appendices D and E provide listing of example Template files and MCNP input files (these are discussed further in Section 4). Appendix F is a complete listing of all of the output data (i.e., all of the analyzed dimensions and

This paper describes a modified version of the MCNP code, the MCNP-DSP. Variance reduction features were disabled to have strictly analog particle tracking in order to follow fluctuating processes more accurately. Some of the neutron and photon physics routines were modified to better represent the production of particles. Other modifications are discussed.

The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX's MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application.

Understanding the limitations of Monte Carlo codes is essential in order to avoid systematic errors in simulations, and to suggest further improvement of the codes. MCNP and EGS4, Monte Carlo codes commonly used in medical physics, were compared and evaluated against electron depth dose data and experimental backscatter results obtained using clinical radiotherapy beams. Different physical models and algorithms used in the codes give significantly different depth dose curves and electron backscattering factors. The default version of MCNP calculates electron depth dose curves which are too penetrating. The MCNP results agree better with experiment if the ITS-style energy-indexing algorithm is used. EGS4 underpredicts electron backscattering for high-Z materials. The results slightly improve if optimal PRESTA-I parameters are used. MCNP simulates backscattering well even for high-Z materials. To conclude the comparison, a timing study was performed. EGS4 is generally faster than MCNP and use of a large number of scoring voxels dramatically slows down the MCNP calculation. However, use of a large number of geometry voxels in MCNP only slightly affects the speed of the calculation.

The purpose of this study was to benchmark the widely used Monte Carlo code MCNP against a set of cold critical experiments with a view to using the code as a means of independently verifying the performance of faster but less accurate Monte Carlo and deterministic codes. The experiments simulated consisted of both fast and thermal criticals as well as fuel in a variety of chemical forms. A standard set of benchmark cold critical experiments was modeled. These included the two fast experiments, GODIVA and JEZEBEL, the TRX metallic uranium thermal experiments, the Babcock and Wilcox oxide and mixed oxidemore » experiments, and the Oak Ridge National Laboratory (ORNL) and Pacific Northwest Laboratory (PNL) nitrate solution experiments. The principal case studied was a small critical experiment that was performed with boiling water reactor bundles.« less

A Theratron 780 (MDS Nordion) 60Co radiotherapy unit has been simulated with the Monte Carlo code MCNP. The unit has been realistically modelled: the cylindrical source capsule and its housing, the rectangular collimator system, both the primary and secondary jaws and the air gaps between the components. Different collimator openings, ranging from 5 x 5 cm2 to 20 x 20 cm2 (narrow and broad beams) at a source-surface distance equal to 80 cm have been used during the study. In the present work, we have calculated spectra as a function of field size. A study of the variation of the electron contamination of the 60Co beam has also been performed.

The transport of electrons and other charged particles is fundamentally different from that of neutrons and photons. A neutron, in aluminum slowing down from 0.5 MeV to 0.0625 MeV will have about 30 collisions; a photon will have fewer than ten. An electron with the same energy loss will undergo 10{sup 5} individual interactions. This great increase in computational complexity makes a single- collision Monte Carlo approach to electron transport unfeasible for many situations of practical interest. Considerable theoretical work has been done to develop a variety of analytic and semi-analytic multiple-scattering theories for the transport of charged particles. Themore » theories used in the algorithms in MCNP are the Goudsmit-Saunderson theory for angular deflections, the Landau an theory of energy-loss fluctuations, and the Blunck-Leisegang enhancements of the Landau theory. In order to follow an electron through a significant energy loss, it is necessary to break the electron`s path into many steps. These steps are chosen to be long enough to encompass many collisions (so that multiple-scattering theories are valid) but short enough that the mean energy loss in any one step is small (for the approximations in the multiple-scattering theories). The energy loss and angular deflection of the electron during each step can then be sampled from probability distributions based on the appropriate multiple- scattering theories. This subsumption of the effects of many individual collisions into single steps that are sampled probabilistically constitutes the ``condensed history`` Monte Carlo method. This method is exemplified in the ETRAN series of electron/photon transport codes. The ETRAN codes are also the basis for the Integrated TIGER Series, a system of general-purpose, application-oriented electron/photon transport codes. The electron physics in MCNP is similar to that of the Integrated TIGER Series.« less

Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

This supporting document contains the software verification and validation (V and V) package used for Phase 2 design of the Spent Nuclear Fuel Multi-Canister Overpack. V and V packages for both ANSYS and MCNP are included. Description of Verification Run(s): This software requires that it be compiled specifically for the machine it is to be used on. Therefore to facilitate ease in the verification process the software automatically runs 25 sample problems to ensure proper installation and compilation. Once the runs are completed the software checks for verification by performing a file comparison on the new output file and themore » old output file. Any differences between any of the files will cause a verification error. Due to the manner in which the verification is completed a verification error does not necessarily indicate a problem. This indicates that a closer look at the output files is needed to determine the cause of the error.« less

The Monte Carlo codes EGS4 and MCNP have been compared when calculating radiotherapy depth doses in water. The aims of the work were to study (i) the differences between calculated depth doses in water for a range of monoenergetic photon energies and (ii) the relative efficiency of the two codes for different electron transport energy cut-offs. The depth doses from the two codes agree with each other within the statistical uncertainties of the calculations (1-2%). The relative depth doses also agree with data tabulated in the British Journal of Radiology Supplement 25. A discrepancy in the dose build-up region may by attributed to the different electron transport algorithims used by EGS4 and MCNP. This discrepancy is considerably reduced when the improved electron transport routines are used in the latest (4B) version of MCNP. Timing calculations show that EGS4 is at least 50% faster than MCNP for the geometries used in the simulations.

The production of accurate detector models is of high importance in the development and use of detectors. Initially, MCNP and Geant were developed to specialise in neutral particle models and accelerator models, respectively; there is now a greater overlap of the capabilities of both, and it is therefore useful to produce comparative models to evaluate detector characteristics. In a collaboration between Lancaster University, UK, and Innovative Physics Ltd., UK, models have been developed in both MCNP6 and Geant4 of Cadmium Zinc Telluride (CdZnTe) detectors developed by Innovative Physics Ltd. Herein, a comparison is made of the relative strengths of MCNP6 and Geant4 for modelling neutron flux and secondary γ-ray emission. Given the increasing overlap of the modelling capabilities of MCNP6 and Geant4, it is worthwhile to comment on differences in results for simulations which have similarities in terms of geometries and source configurations.

A patch to MCNP5 has been written to allow generation of multiple neutrons from a spontaneous-fission event and generate list-mode output. This report documents the implementation and usage of this patch.

The representative point method for the efficiency calibration of volume samples has been previously proposed. For smoothly implementing the method, a calculation code named CREPT-MCNP has been developed. The code estimates the position of a representative point which is intrinsic to each shape of volume sample. The self-absorption correction factors are also given to make correction on the efficiencies measured at the representative point with a standard point source. Features of the CREPT-MCNP code are presented.

MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 usingmore » CEM03.03 and LAQGSM03.03 describes well fragmentation reactions induced on light and medium target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below, and can serve as a reliable simulation tool for different applications, like cosmic-ray-induced single event upsets (SEU’s), radiation protection, and cancer therapy with proton and ion beams, to name just a few. Future improvements of the predicting capabilities of MCNP6 for such reactions are possible, and are discussed in this work.« less

These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such asmore » MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.« less

We have focused on the usage of MCNP code for calculation of Gamma Knife radiation field parameters with a homogenous polystyrene phantom. We have investigated several parameters of the Leksell Gamma Knife radiation field and compared the results with other studies based on EGS4 and PENELOPE code as well as the Leksell Gamma Knife treatment planning system Leksell GammaPlan (LGP). The current model describes all 201 radiation beams together and simulates all the sources in the same time. Within each beam, it considers the technical construction of the source, the source holder, collimator system, the spherical phantom, and surrounding material. We have calculated output factors for various sizes of scoring volumes, relative dose distributions along basic planes including linear dose profiles, integral doses in various volumes, and differential dose volume histograms. All the parameters have been calculated for each collimator size and for the isocentric configuration of the phantom. We have found the calculated output factors to be in agreement with other authors' works except the case of 4 mm collimator size, where averaging over the scoring volume and statistical uncertainties strongly influences the calculated results. In general, all the results are dependent on the choice of the scoring volume. The calculated linear dose profiles and relative dose distributions also match independent studies and the Leksell GammaPlan, but care must be taken about the fluctuations within the plateau, which can influence the normalization, and accuracy in determining the isocenter position, which is important for comparing different dose profiles. The calculated differential dose volume histograms and integral doses have been compared with data provided by the Leksell GammaPlan. The dose volume histograms are in good agreement as well as integral doses calculated in small calculation matrix volumes. However, deviations in integral doses up to 50% can be observed for large

This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.

Ion Chambers have a generally flat energy response with some deviations at very low (<100 keV) and very high (>2 MeV) energies. Some improvements in the low energy response can be achieved through use of high atomic number gases, such as argon and xenon, and higher chamber pressures. This work looks at the energy response of high pressure xenon-filled ion chambers using the MCNP Monte Carlo package to develop geometric models of a commercially available high pressure ion chamber (HPIC). The use of the F6 tally as an estimator of the energy deposited in a region of interest per unit mass, and the underlying assumptions associated with its use are described. The effect of gas composition, chamber gas pressure, chamber wall thickness, and chamber holder wall thicknesses on energy response are investigated and reported. The predicted energy response curve for the HPIC was found to be similar to that reported by other investigators. These investigations indicate that improvements to flatten the overall energy response of the HPIC down to 70 keV could be achieved through use of 3 mm-thick stainless steel walls for the ion chamber.

While neutron activation analysis is widely used in many areas, sensitivity of the analysis depends on how the analysis is conducted. Even though the sensitivity of the techniques carries error, compared to chemical analysis, its range is in parts per million or sometimes billion. Due to this sensitivity, the use of neutron activation analysis becomes important when analyzing bio-samples. Artificial neural network is an attractive technique for complex systems. Although there are neural network applications on spectral analysis, training by simulated data to analyze experimental data has not been made. This study offers an improvement on spectral analysis and optimization on neural network for the purpose. The work considers five elements that are considered as trace elements for bio-samples. However, the system is not limited to five elements. The only limitation of the study comes from data library availability on MCNP. A perceptron network was employed to identify five elements from gamma spectra. In quantitative analysis, better results were obtained when the neural fitting tool in MATLAB was used. As a training function, Levenberg-Marquardt algorithm was used with 23 neurons in the hidden layer with 259 gamma spectra in the input. Because the interest of the study deals with five elements, five neurons representing peak counts of five isotopes in the input layer were used. Five output neurons revealed mass information of these elements from irradiated kidney stones. Results showing max error of 17.9% in APA, 24.9% in UA, 28.2% in COM, 27.9% in STRU type showed the success of neural network approach in analyzing gamma spectra. This high error was attributed to Zn that has a very long decay half-life compared to the other elements. The simulation and experiments were made under certain experimental setup (3 hours irradiation, 96 hours decay time, 8 hours counting time). Nevertheless, the approach is subject to be generalized for different setups.

The Monte Carlo computer code MCNP (version 4A) has been used to develop a personal computer-based model of the Swansea in vivo neutron activation analysis (IVNAA) system. The model included specification of the neutron source (252Cf), collimators, reflectors and shielding. The MCNP model was 'benchmarked' against fast neutron and thermal neutron fluence data obtained experimentally from the IVNAA system. The Swansea system allows two irradiation geometries using 'short' and 'long' collimators, which provide alternative dose rates for IVNAA. The data presented here relate to the short collimator, although results of similar accuracy were obtained using the long collimator. The fast neutron fluence was measured in air at a series of depths inside the collimator. The measurements agreed with the MCNP simulation within the statistical uncertainty (5-10%) of the calculations. The thermal neutron fluence was measured and calculated inside the cuboidal water phantom. The depth of maximum thermal fluence was 3.2 cm (measured) and 3.0 cm (calculated). The width of the 50% thermal fluence level across the phantom at its mid-depth was found to be the same by both MCNP and experiment. This benchmarking exercise has given us a high degree of confidence in MCNP as a tool for the design of IVNAA systems.

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the pointmore » reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.« less

Most of the development of the MCNPX/6 burnup capability focused on features that were applied to the Boltzman transport or used to prepare coefficients for use in CINDER90, with little change to CINDER90 or the CINDER90 data. Though a scheme exists for best solving the coupled Boltzman and Bateman equations, the most significant approximation is that the employed nuclear data are correct and complete. Thus, the CINDER90 library file contains 60 different actinide fission yields encompassing 36 fissionable actinides (thermal, fast, high energy and spontaneous fission). Fission reaction data exists for more than 60 actinides and as a result, fissionmore » yield data must be approximated for actinides that do not possess fission yield information. Several types of approximations are used for estimating fission yields for actinides which do not possess explicit fission yield data. The objective of this study is to test whether or not certain approximations of fission yield selection have any impact on predictability of major actinides and fission products. Further we assess which other fission products, available in MCNP6 Tier 3, result in the largest difference in production. Because the CINDER90 library file is in ASCII format and therefore easily amendable, we assess reasons for choosing, as well as compare actinide and major fission product prediction for the H. B. Robinson benchmark for, three separate fission yield selection methods: (1) the current CINDER90 library file method (Base); (2) the element method (Element); and (3) the isobar method (Isobar). Results show that the three methods tested result in similar prediction of major actinides, Tc-99 and Cs-137; however, certain fission products resulted in significantly different production depending on the method of choice.« less

The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a selectmore » group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability

The calculation of the thermal neutron Doppler temperature reactivity feedback co-efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective temperature derivatives within the system in order to accurately calculate the Doppler temperature feedback. A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) generated continuous energy cross sections within the Monte Carlo N-Particle (MCNP6) transport code. The adjoint-weighted tallies are generated during the continuous energy k-eigenvalue Monte Carlo calculation. The weighting is based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. The adjoint-weighted tallies are produced in a forward calculation and do not require an inversion of the random walk. The OTF cross section database uses a high order functional expansion between points on a user-defined energy-temperature mesh in which the coefficients with respect to a polynomial fitting in temperature are stored. The coefficients of the fits are generated before run- time and called upon during the simulation to produce cross sections at any given energy and temperature. The polynomial form of the OTF cross sections allows the possibility of obtaining temperature derivatives of the cross sections on-the-fly. The use of Monte Carlo sampling of adjoint-weighted tallies and the capability of computing derivatives of continuous energy cross sections with respect to temperature are used to calculate the Doppler temperature coefficient in a research

The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators.

Interface software was developed to generate the input file to run Monte Carlo MCNP-4B code from medical image in Interfile format version 3.3. The software was tested using a spherical phantom of tomography slides with known cumulated activity distribution in Interfile format generated with IMAGAMMA medical image processing system. The 3D dose calculation obtained with Monte Carlo MCNP-4B code was compared with the voxel S factor method. The results show a relative error between both methods less than 1 %.

In this paper, an overview of the Swiss Nuclear Power Plant (NPP) activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG) in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

energy ranges. Accordingly, special care has to be taken in setting choice for calculating electron dose distribution with MCNP6, in particular with regards to dosimetry or nuclear medicine applications.

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.

The present work has simulated the photon and electron transport in a Theratron 780® (MDS Nordion)60Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle). This project explains mainly the different methodologies carried out to speedup calculations in order to apply this code efficiently in radiotherapy treatment planning.

The performance of seven computer platforms is evaluated with the widely used and internationally available MCNP4A Monte Carlo radiation transport code. All results are reproducible and are presented in such a way as to enable comparison with computer platforms not in the study. The authors observed that the HP/9000-735 workstation runs MCNP 50% faster than the Cray YMP 8/64. Compared with the Cray YMP 8/64, the IBM RS/6000-560 is 68% as fast, the Sun Sparc10 is 66% as fast, the Silicon Graphics ONYX is 90% as fast, the Gateway 2000 model 4DX2-66V personal computer is 27% as fast, and themore » Sun Sparc2 is 24% as fast. In addition to comparing the timing performance of the seven platforms, the authors observe that changes in compilers and software over the past 2 yr have resulted in only modest performance improvements, hardware improvements have enhanced performance by less than a factor of [approximately]3, timing studies are very problem dependent, MCNP4Q runs about as fast as MCNP4.« less

The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (k eff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the biasmore » due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.« less

The linear attenuation coefficient, mass attenuation coefficient and mean free path of various Lead-Boron Polyethylene (PbBPE) samples which can be used as the photon shielding materials in marine reactor have been simulated using the Monte Carlo N-Particle (MCNP)-5 code. The MCNP simulation results are in good agreement with the XCOM values and the reported experimental data for source Cesium-137 and Cobalt-60. Thus, this method based on MCNP can be used to simulate the photon attenuation characteristics of various types of PbBPE materials.

The dose distributions from proton pencil beam scanning were calculated by FLUKA, GEANT4, MCNP, and PHITS, in order to investigate their applicability in proton radiotherapy. The first studied case was the integrated depth dose curves (IDDCs), respectively from a 100 and a 226-MeV proton pencil beam impinging a water phantom. The calculated IDDCs agree with each other as long as each code employs 75 eV for the ionization potential of water. The second case considered a similar condition of the first case but with proton energies in a Gaussian distribution. The comparison to the measurement indicates the inter-code differences might not only due to different stopping power but also the nuclear physics models. How the physics parameter setting affect the computation time was also discussed. In the third case, the applicability of each code for pencil beam scanning was confirmed by delivering a uniform volumetric dose distribution based on the treatment plan, and the results showed general agreement between each codes, the treatment plan, and the measurement, except that some deviations were found in the penumbra region. This study has demonstrated that the selected codes are all capable of performing dose calculations for therapeutic scanning proton beams with proper physics settings.

In the nuclear engineering, you have to manage time and precision. Especially in shielding design, you have to be more accurate and efficient to reduce cost (shielding thickness optimization), and for this, you use 3D codes. In this paper, we want to see if we can easily applicate the CADIS methods for design shielding of small pipes which go through large concrete walls. We assess the impact of the WW generated by the 3D-deterministic code ATTILA versus WW directly generated by MCNP (iterative and manual process). The comparison is based on the quality of the convergence (estimated relative error (σ), Variance of Variance (VOV) and Figure of Merit (FOM)), on time (computer time + modelling) and on the implement for the engineer.

The Taylor series or differential operator perturbation method, implemented in MCNP and invoked using the PERT card, can be used for efficient parameter studies in shielding problems. This paper shows how only two PERT cards are needed to generate an entire parameter study, including statistical uncertainty estimates (an additional three PERT cards can be used to give exact statistical uncertainties). One realistic example problem involves a detailed helium-3 neutron detector model and its efficiency as a function of the density of its high-density polyethylene moderator. The MCNP differential operator perturbation capability is extremely accurate for this problem. A second problem involves the density of the polyethylene reflector of the BeRP ball and is an example of first-order sensitivity analysis using the PERT capability. A third problem is an analytic verification of the PERT capability.

A new approach for calculating internal dose estimates was developed through the use of a more realistic computational model of the human body. The present technique shows the capability to build a patient-specific phantom with tomography data (a voxel-based phantom) for the simulation of radiation transport and energy deposition using Monte Carlo methods such as in the MCNP-4B code. MCNP-4B absorbed fractions for photons in the mathematical phantom of Snyder et al. agreed well with reference values. Results obtained through radiation transport simulation in the voxel-based phantom, in general, agreed well with reference values. Considerable discrepancies, however, were found in some cases due to two major causes: differences in the organ masses between the phantoms and the occurrence of organ overlap in the voxel-based phantom, which is not considered in the mathematical phantom.

A Small Angle Neutron Scattering instrument is being installed at one end of the tangential beam tube of the Portuguese Research Reactor. The instrument is fed using a neutron scatterer positioned in the middle of the beam tube. The scatterer consists of circulating H2O contained in a hollow disc of Al. The in-pile shielding components and the shielding installed around the neutron selector have been the object of an MCNP simulation study. The quantities calculated were the neutron and gamma-ray fluxes in different positions, the energy deposited in the material by the neutron and gamma-ray fields, the material activation resulting from the neutron field and radiation doses at the exit wall of the shutter and around the shielding. The MCNP results are presented and compared with results of an analytical approach and with experimental data collected after installation.

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

gamma-ray spectra of natural radionuclides are simulated for a BGO detector in a borehole geometry using the Monte Carlo code MCNP. All gamma-ray emissions of the decay of 40K and the series of 232Th and 238U are used to describe the source. A procedure is proposed which excludes the time-consuming electron tracking in less relevant areas of the geometry. The simulated gamma-ray spectra are benchmarked against laboratory data.

The paper presents results from MCNP6 simulations of galactic cosmic ray (GCR) propagation down through the Martian atmosphere to the surface and comparison with RAD measurements made there. This effort is part of a collaborative modeling workshop for space radiation hosted by Southwest Research Institute (SwRI). All modeling teams were tasked with simulating the galactic cosmic ray (GCR) spectrum through the Martian atmosphere and the Radiation Assessment Detector (RAD) on-board the Curiosity rover. The detector had two separate particle acceptance angles, 4π and 30 ° off zenith. All ions with Z = 1 through Z = 28 were tracked in both scenarios while some additional secondary particles were only tracked in the 4π cases. The MCNP6 4π absorbed dose rate was 307.3 ± 1.3 μGy/day while RAD measured 233 μGy/day. Using the ICRP-60 dose equivalent conversion factors built into MCNP6, the simulated 4π dose equivalent rate was found to be 473.1 ± 2.4 μSv/day while RAD reported 710 μSv/day.

Monte Carlo calculations are frequently used to analyse a variety of radiological science applications using low-energy (10-1000 keV) photon sources. This study seeks to create a low-energy benchmark for the MCNP Monte Carlo code by simulating the absolute dose rate in water and the air-kerma rate for monoenergetic point sources with energies between 10 keV and 1 MeV. The analysis compares four cross-section datasets as well as the tally method for collision kerma versus absorbed dose. The total photon attenuation coefficient cross-section for low atomic number elements has changed significantly as cross-section data have changed between 1967 and 1989. Differences of up to 10% are observed in the photoelectric cross-section for water at 30 keV between the standard MCNP cross-section dataset (DLC-200) and the most recent XCOM/NIST tabulation. At 30 keV, the absolute dose rate in water at 1.0 cm from the source increases by 7.8% after replacing the DLC-200 photoelectric cross-sections for water with those from the XCOM/NIST tabulation. The differences in the absolute dose rate are analysed when calculated with either the MCNP absorbed dose tally or the collision kerma tally. Significant differences between the collision kerma tally and the absorbed dose tally can occur when using the DLC-200 attenuation coefficients in conjunction with a modern tabulation of mass energy-absorption coefficients.

As part of an ongoing dose reconstruction project, equivalent organ dose rates from photons and neutrons were estimated using the energy spectra measured in the central hall above the graphite reactor core located in the Russian Mayak Production Association facility. Reconstruction of the work environment was necessary due to the lack of personal dosimeter data for neutrons in the time period prior to 1987. A typical worker scenario for the central hall was developed for the Monte Carlo Neutron Photon-4B (MCNP) code. The resultant equivalent dose rates for neutrons and photons were compared with the equivalent dose rates derived from calculations using the conversion coefficients in the International Commission on Radiological Protection Publications 51 and 74 in order to validate the model scenario for this Russian facility. The MCNP results were in good agreement with the results of the ICRP publications indicating the modeling scenario was consistent with actual work conditions given the spectra provided. The MCNP code will allow for additional orientations to accurately reflect source locations.

Voxelized geometry such as those obtained from medical images is increasingly used in Monte Carlo calculations of absorbed doses. One useful application of calculated absorbed dose is the determination of fluence-to-dose conversion factors for different organs. However, confusion still exists about how such a geometry is defined and how the energy deposition is best computed, especially involving a popular code, MCNP5. This study investigated two different types of geometries in the MCNP5 code, cell and lattice definitions. A 10 cm x 10 cm x 10 cm test phantom, which contained an embedded 2 cm x 2 cm x 2 cm target at its center, was considered. A planar source emitting parallel photons was also considered in the study. The results revealed that MCNP5 does not calculate total target volume for multi-voxel geometries. Therefore, tallies which involve total target volume must be divided by the user by the total number of voxels to obtain a correct dose result. Also, using planar source areas greater than the phantom size results in the same fluence-to-dose conversion factor.

Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) Pool Critical Assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varyingmore » ages (those packaged with MCNP6, from the IRDF-2002 multi-group library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction- and dosimetry location-basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but is typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.« less

AWE performs experiments to investigate the hydrodynamic behavior of explosive metal systems in order to underwrite the UK nuclear deterrent. The experiments involve the manufacture of a device to mock up some aspect of the weapon. Inert simulant materials replace fissile weapon components. The device is then detonated under remote control within specially designed explosive containment buildings called firing chambers. During the experiment a very brief, intense, collimated flash of high energy x-rays are used to take a snapshot of the implosion (see Fig. 1). Prom the resulting image measurements of the dynamic configuration and density distribution of the components in the device are inferred. These are then used to compare with calculations of the hydrodynamic operation of the weapon and understand how the device would perform under various conditions. This type of experiment is known as a core punch experiment.

This report presents a set of validation and verification (V&V) MCNP 6.1.1 beta results calculated in parallel, with MPI, obtained using its event generators at intermediate and high-energies compared against various experimental data. It also contains several examples of results using the models at energies below 150 MeV, down to 10 MeV, where data libraries are normally used. This report can be considered as the forth part of a set of MCNP6 Testing Primers, after its first, LA-UR-11-05129, and second, LA-UR-11-05627, and third, LA-UR-26944, publications, but is devoted to V&V with the latest, 1.1 beta version of MCNP6. The MCNP6more » test-problems discussed here are presented in the /VALIDATION_CEM/and/VALIDATION_LAQGSM/subdirectories in the MCNP6/Testing/directory. README files that contain short descriptions of every input file, the experiment, the quantity of interest that the experiment measures and its description in the MCNP6 output files, and the publication reference of that experiment are presented for every test problem. Templates for plotting the corresponding results with xmgrace as well as pdf files with figures representing the final results of our V&V efforts are presented. Several technical “bugs” in MCNP 6.1.1 beta were discovered during our current V&V of MCNP6 while running it in parallel with MPI using its event generators. These “bugs” are to be fixed in the following version of MCNP6. Our results show that MCNP 6.1.1 beta using its CEM03.03, LAQGSM03.03, Bertini, and INCL+ABLA, event generators describes, as a rule, reasonably well different intermediate- and high-energy measured data. This primer isn’t meant to be read from cover to cover. Readers may skip some sections and go directly to any test problem in which they are interested.« less

In this paper, a detailed study of the electron transport in MCNP is performed, separating the effects of the energy binning technique on the energy loss rate, the scattering angles, and the sub-step length as a function of energy. As this problem is already well known, in this paper we focus on the explanation as to why the default mode of MCNP can lead to large deviations. The resolution dependence was investigated as well. An error in the MCNP code in the energy binning technique in the default mode (DBCN 18 card = 0) was revealed, more specific in the updating of cross sections when a sub-step is performed corresponding to a high-energy loss. This updating error is not present in the ITS mode (DBCN 18 card = 1) and leads to a systematically lower dose deposition rate in the default mode. The effect is present for all energies studied (0.5-10 MeV) and depends on the geometrical resolution of the scoring regions and the energy grid resolution. The effect of the energy binning technique is of the same order of that of the updating error for energies below 2 MeV, and becomes less important for higher energies. For a 1 MeV point source surrounded by homogeneous water, the deviation of the default MCNP results at short distances attains 9% and remains approximately the same for all energies. This effect could be corrected by removing the completion of an energy step each time an electron changes from an energy bin during a sub-step. Another solution consists of performing all calculations in the ITS mode. Another problem is the resolution dependence, even in the ITS mode. The higher the resolution is chosen (the smaller the scoring regions) the faster the energy is deposited along the electron track. It is proven that this is caused by starting a new energy step when crossing a surface. The resolution effect should be investigated for every specific case when calculating dose distributions around beta sources. The resolution should not be higher than 0.85*(1-EFAC

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

The Forschungszentrum Juelich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of somemore » elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.« less

The Forschungszentrum Jülich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA) [1]. The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.

In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure-vessel design with 451 vertical coolant channels, and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update calculation methods and models (cell, supercell and reactor) was recently carried out coveringmore » cell, supercell (control rod) and core calculations. As a validation of the new models some benchmark comparisons were done with Monte Carlo calculations with MCNP5. This paper presents comparisons of cell and supercell benchmark problems based on a slightly idealized model of the Atucha-I core obtained with the WIMS-D5 and DRAGON codes with MCNP5 results. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, and more symmetric than Atucha-II Cell parameters compared include cell k-infinity, relative power levels of the different rings of fuel rods, and some two-group macroscopic cross sections. Supercell comparisons include supercell k-infinity changes due to the control rods (tubes) of steel and hafnium. (authors)« less

The feasibility of modeling a pressurized-water-reactor (PWR) facility and calculating dose rates at all locations within the containment and adjoining structures using MCNP5 with mesh tallies is presented. Calculations of dose rates resulting from neutron and photon sources from the reactor (operating and shut down for various periods) and the spent fuel pool, as well as for the photon source from the primary coolant loop, were all of interest. Identification of the PWR facility, development of the MCNP-based model and automation of the run process, calculation of the various sources, and development of methods for visually examining mesh tally filesmore » and extracting dose rates were all a significant part of the project. Advanced variance reduction, which was required because of the size of the model and the large amount of shielding, was performed via the CADIS/ADVANTG approach. This methodology uses an automatically generated three-dimensional discrete ordinates model to calculate adjoint fluxes from which MCNP weight windows and source bias parameters are generated. Investigative calculations were performed using a simple block model and a simplified full-scale model of the PWR containment, in which the adjoint source was placed in various regions. In general, it was shown that placement of the adjoint source on the periphery of the model provided adequate results for regions reasonably close to the source (e.g., within the containment structure for the reactor source). A modification to the CADIS/ADVANTG methodology was also studied in which a global adjoint source is weighted by the reciprocal of the dose response calculated by an earlier forward discrete ordinates calculation. This method showed improved results over those using the standard CADIS/ADVANTG approach, and its further investigation is recommended for future efforts.« less

SABRINA is an interactive three-dimensional geometry modeler developed to produce complicated models for the Los Alamos Monte Carlo Neutron Photon program MCNP. SABRINA produces line drawings and color-shaded drawings for a wide variety of interactive graphics terminals. It is used as a geometry preprocessor in model development and as a Monte Carlo particle-track postprocessor in the visualization of complicated particle transport problem. SABRINA is written in Fortran 77 and is based on the Los Alamos Common Graphics System, CGS. 5 refs., 2 figs.

The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.

Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple_ace.pl and simple_ace_mg.pl.

The efficiency of a whole-body counter for (137)Cs and (40)K was calculated using the MCNP5 code. The ORNL phantoms of a human body of different body sizes were applied in a sitting position in front of a detector. The aim was to investigate the dependence of efficiency on the body size (age) and the detector position with respect to the body and to estimate the accuracy of real measurements. The calculation work presented here is related to the NaI detector, which is available in the Serbian Whole-body Counter facility in Vinca Institute.

By using MCNP code and ethanol-chlorobenzene (ECB) dosimeters the simulations and measurements of absorbed dose distribution in a tote-box of the Cobalt-60 irradiator, SVST-Co60/B at VINAGAMMA have been done. Based on the results Dose Uniformity Ratios (DUR), positions and values of minimum and maximum dose extremes in a tote-box, and efficiency of the irradiator for the different dummy densities have been gained. There is a good agreement between simulation and experimental results in comparison and they have valuable meanings for operation of the irradiator. Copyright 2010 Elsevier Ltd. All rights reserved.

Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this work. Calculations have been done for the most common natural radionuclides in soil (238)U, (232)Th series and (40)K. A ORNL human phantoms and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs. The effective dose was calculated according to ICRP 74 recommendations. Conversion factors of effective dose per air kerma were determined. Results obtained here were compared with other authors. Copyright 2009 Elsevier Ltd. All rights reserved.

The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operatingmore » scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of

Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

64 Cu emits positrons as well as β - particles and Auger and internal conversion electrons useful for radiotherapy. Our objective was to model the cellular dosimetry of 64 Cu under different geometries commonly used to study the cytotoxic effects of 64 Cu. Monte Carlo N-Particle (MCNP) was used to simulate the transport of all particles emitted by 64 Cu from the cell surface (CS), cytoplasm (Cy), or nucleus (N) of a single cell; monolayer in a well (radius = 0.32-1.74 cm); or a sphere (radius = 50-6,000 μm) of cells to calculate S values. The radius of the cell and N ranged from 5 to 12 μm and 2 to 11 μm, respectively. S values were obtained by MIRDcell for comparison. MCF7/HER2-18 cells were exposed in vitro to 64 Cu-labeled trastuzumab. The subcellular distribution of 64 Cu was measured by cell fractionation. The surviving fraction was determined in a clonogenic assay. The relative differences of MCNP versus MIRDcell self-dose S values (S self ) for 64 Cu ranged from -0.2% to 3.6% for N to N (S N←N ), 2.3% to 8.6% for Cy to N (S N←Cy ), and -12.0% to 7.3% for CS to N (S N←CS ). The relative differences of MCNP versus MIRDcell cross-dose S values were 25.8%-30.6% for a monolayer and 30%-34% for a sphere, respectively. The ratios of S N←N versus S N←Cy and S N←Cy versus S N←CS decreased with increasing ratio of the N of the cell versus radius of the cell and the size of the monolayer or sphere. The surviving fraction of MCF7 /: HER2-18 cells treated with 64 Cu-labeled trastuzumab (0.016-0.368 MBq/μg, 67 nM) for 18 h versus the absorbed dose followed a linear survival curve with α = 0.51 ± 0.05 Gy -1 and R 2 = 0.8838. This is significantly different from the linear quadratic survival curve of MCF7 /: HER2-18 cells exposed to γ-rays. MCNP- and MIRDcell-calculated S values agreed well. 64 Cu in the N increases the dose to the N in isolated single cells but has less effect in a cell monolayer or small cluster of cells simulating a micrometastasis

Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross‐sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross‐sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code — MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low‐energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes. PACS number(s): 87.56.bg PMID:27074460

Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in 125I and 103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as 125I and 103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for 103Pd and 10 cm for 125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for 192Ir and less than 1.2% for 137Cs between the three codes.

This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using "TORT-coupled MCNP" method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency.

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (50 MeV to 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used inmore » the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are now available. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results.« less

Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient’s body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5. PMID:28362837

Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (50 MeV to 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used inmore » the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are now available. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results.« less

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.

Tissue-equivalent proportional counters (TEPCs) utilise tissue-equivalent materials to depict homogeneous microscopic volumes of human tissue. Although both the walls and gas simulate the same medium, they respond to radiation differently. Density differences between the two materials cause distortions, or wall effects, in measurements, with the most dominant effect caused by delta rays. This study uses a Monte Carlo transport code, MCNP, to simulate the transport of secondary electrons within a TEPC. The Rudd model, a singly differential cross section with no dependence on electron direction, is used to describe the energy spectrum obtained by the impact of two iron beams on water. Based on the models used in this study, a wall-less TEPC had a higher lineal energy (keV.micron-1) as a function of impact parameter than a solid-wall TEPC for the iron beams under consideration. An important conclusion of this study is that MCNP has the ability to model the wall effects observed in TEPCs.

Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient's body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5.

This report documents the work done by X-6, HSE-6, and CTR Technical Services to produce a comprehensive working cross-section library for MCNP-4 suitable for SUN workstations and similar environments. The resulting library consists of a total of 436 files (one file for each ZAID). The library is 152 Megabytes in Type 1 format and 32 Megabytes in Type 2 format. Type 2 can be used when porting the library from one computer to another of the same make. Otherwise, Type 1 must be used to ensure portability between different computer systems. Instructions for installing the library and adding ZAIDs tomore » it are included here. Also included is a description of the steps necessary to install and test version 4 of MCNP. To improve readability of this report, certain commands and filenames are given in uppercase letters. The actual command or filename on the SUN workstation, however, must be specified in lowercase letters. Any questions regarding the data contained in the library should be directed to X-6 and any questions regarding the installation of the library and the testing that was performed should be directed to HSE-6. 9 refs., 7 tabs.« less

The Boron Neutron Capture Therapy ( BNCT) is based on selective uptake of boron in tumour tissue compared to the surrounding normal tissue. Infusion of compounds with boron is followed by irradiation with neutrons. Neutron capture on 10B, which gives rise to an alpha particle and recoiled 7Li ion, enables the therapeutic dose to be delivered to tumour tissue while healthy tissue can be spared. Here, therapeutic abilities of BNCT were studied for possible treatment of liver cancer using thermal and epithermal neutron beam. For neutron transport MCNP software was used and doses in organs of interest in ORNL phantom were evaluated. Phantom organs were filled with voxels in order to obtain depth-dose distributions in them. The result suggests that BNCT using an epithermal neutron beam could be applied for liver cancer treatment.

Wmore » hen multiple neutrons are emitted in a fission event they are correlated in both energy and their relative angle, which may impact the design of safeguards equipment and other instrumentation for non-proliferation applications. The most recent release of MCNP 6 . 2 contains the capability to simulate correlated fission neutrons using the event generators CGMF and FREYA . These radiation transport simulations will be post-processed by the detector response code, DRiFT , and compared directly to correlated fission measurements. DRiFT has been previously compared to single detector measurements, its capabilities have been recently expanded with correlated fission simulations in mind. Finally, this paper details updates to DRiFT specific to correlated fission measurements, including tracking source particle energy of all detector events (and non-events), expanded output formats, and digitizer waveform generation.« less

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution).more » Comparison between both solutions is briefly illustrated in this summary.« less

Wmore » hen multiple neutrons are emitted in a fission event they are correlated in both energy and their relative angle, which may impact the design of safeguards equipment and other instrumentation for non-proliferation applications. The most recent release of MCNP 6 . 2 contains the capability to simulate correlated fission neutrons using the event generators CGMF and FREYA . These radiation transport simulations will be post-processed by the detector response code, DRiFT , and compared directly to correlated fission measurements. DRiFT has been previously compared to single detector measurements, its capabilities have been recently expanded with correlated fission simulations in mind. Finally, this paper details updates to DRiFT specific to correlated fission measurements, including tracking source particle energy of all detector events (and non-events), expanded output formats, and digitizer waveform generation.« less

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

In this work, the MCNP4B code has been employed to calculate conversion coefficients from air kerma to the ambient dose equivalent, H*(10)/Ka, for monoenergetic photon energies from 10 keV to 50 MeV, assuming the kerma approximation. Also estimated are the H*(10)/Ka for photon beams produced by linear accelerators, such as Clinac-4 and Clinac-2500, after transmission through primary barriers of radiotherapy treatment rooms. The results for the conversion coefficients for monoenergetic photon energies, with statistical uncertainty <2%, are compared with those in ICRP publication 74 and good agreements were obtained. The conversion coefficients calculated for real clinic spectra transmitted through walls of concrete of 1, 1.5 and 2 m thick, are in the range of 1.06-1.12 Sv Gy(-1).

A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.

Absorbed doses caused by electron irradiation were calculated with Monte Carlo N-Particle transport code (MCNP) for thin layers of tissue-equivalent materials. The layers were so thin that the calculation of energy deposition was on the border of the scope of MCNP. Therefore, in this article application of three different methods of calculation of energy deposition is discussed. This was done by means of two scenarios: in the first one, electrons were emitted from the centre of a sphere of water and also recorded in that sphere; and in the second, an irradiation with the PTB Secondary Standard BSS2 was modelled, where electrons were emitted from an (90)Sr/(90)Y area source and recorded inside a cuboid phantom made of tissue-equivalent material. The speed and accuracy of the different methods were of interest. While a significant difference in accuracy was visible for one method in the first scenario, the difference in accuracy of the three methods was insignificant for the second one. Considerable differences in speed were found for both scenarios. In order to demonstrate the need for calculating the dose in thin small zones, a third scenario was constructed and simulated as well. The third scenario was nearly equal to the second one, but a pike of lead was assumed to be inside the phantom in addition. A dose enhancement (caused by the pike of lead) of ∼113 % was recorded for a thin hollow cylinder at a depth of 0.007 cm, which the basal-skin layer is referred to in particular. Dose enhancements between 68 and 88 % were found for a slab with a radius of 0.09 cm for all depths. All dose enhancements were hardly noticeable for a slab with a cross-sectional area of 1 cm(2), which is usually applied to operational radiation protection.

Boron Neutron Capture Therapy (BNCT) is used for treatment of many diseases, including brain tumors, in many medical centers. In this method, a target area (e.g., head of patient) is irradiated by some optimized and suitable neutron fields such as research nuclear reactors. Aiming at protection of healthy tissues which are located in the vicinity of irradiated tissue, and based on the ALARA principle, it is required to prevent unnecessary exposure of these vital organs. In this study, by using numerical simulation method (MCNP4C Code), the absorbed dose in target tissue and the equiavalent dose in different sensitive tissues of a patiant treated by BNCT, are calculated. For this purpose, we have used the parameters of MIRD Standard Phantom. Equiavelent dose in 11 sensitive organs, located in the vicinity of target, and total equivalent dose in whole body, have been calculated. The results show that the absorbed dose in tumor and normal tissue of brain equal to 30.35 Gy and 0.19 Gy, respectively. Also, total equivalent dose in 11 sensitive organs, other than tumor and normal tissue of brain, is equal to 14 mGy. The maximum equivalent doses in organs, other than brain and tumor, appear to the tissues of lungs and thyroid and are equal to 7.35 mSv and 3.00 mSv, respectively.

The Coolant Void Reactivity (CVR) is an important safety parameter that needs to be estimated at the design stage of a nuclear reactor. It helps to have an a priori knowledge of the behavior of the system during a transient initiated by the loss of coolant. In the present paper, we have attempted to estimate the CVR for a CANDU New Generation (CANDU-NG) lattice, as proposed at an early stage of the Advanced CANDU Reactor (ACR) development. We have attempted to estimate the CVR with development version of the code DRAGON, using the method of characteristics. DRAGON has several advancedmore » self-shielding models incorporated in it, each of them compatible with the method of characteristics. This study will bring to focus the performance of these self-shielding models, especially when there is voiding of such a tight lattice. We have also performed assembly calculations in 2 x 2 pattern for the CANDU-NG fuel, with special emphasis on checkerboard voiding. The results obtained have been validated against Monte Carlo codes MCNP5 and TRIPOLI-4.3. (authors)« less

Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

Three Monte Carlo photon/electron transport codes (GEPTS, EGSnrc, and MCNP) are bench-marked against dose measurements in homogeneous (both low- and high-Z) media as well as at interfaces. A brief overview on physical models used by each code for photon and electron (positron) transport is given. Absolute calorimetric dose measurements for 0.5 and 1 MeV electron beams incident on homogeneous and multilayer media are compared with the predictions of the three codes. Comparison with dose measurements in two-layer media exposed to a 60Co gamma source is also performed. In addition, comparisons between the codes (including the EGS4 code) are done for (a) 0.05 to 10 MeV electron beams and positron point sources in lead, (b) high-energy photons (10 and 20 MeV) irradiating a multilayer phantom (water/steel/air), and (c) simulation of a 90Sr/90Y brachytherapy source. A good agreement is observed between the calorimetric electron dose measurements and predictions of GEPTS and EGSnrc in both homogeneous and multilayer media. MCNP outputs are found to be dependent on the energy-indexing method (Default/ITS style). This dependence is significant in homogeneous media as well as at interfaces. MCNP(ITS) fits more closely the experimental data than MCNP(DEF), except for the case of Be. At low energy (0.05 and 0.1 MeV), MCNP(ITS) dose distributions in lead show higher maximums in comparison with GEPTS and EGSnrc. EGS4 produces too penetrating electron-dose distributions in high-Z media, especially at low energy (<0.1 MeV). For positrons, differences between GEPTS and EGSnrc are observed in lead because GEPTS distinguishes positrons from electrons for both elastic multiple scattering and bremsstrahlung emission models. For the 60Co source, a quite good agreement between calculations and measurements is observed with regards to the experimental uncertainty. For the other cases (10 and 20 MeV photon sources and the 90Sr/90Y beta source), a good agreement is found between the three

The relative response of the new Health Protection Agency thermoluminescence dosimeter (TLD) has been calculated for Narrow Series X-ray distribution and (137)Cs photon sources using the Monte Carlo code MCNP5, and the results compared with those obtained during its design stage using the predecessor code, MCNP4c2. The results agreed at intermediate energies (approximately 0.1 MeV to (137)Cs), but differed at low energies (<0.1 MeV) by up to approximately 10%. This disparity has been ascribed to differences in the default photon interaction data used by the two codes, and derives ultimately from the effect on absorbed dose of the recent updates to the photoelectric cross sections. The sources of these data have been reviewed.

Emerging radiological applications of gold nanoparticles demand low-energy electron/photon transport calculations including details of an atomic relaxation process. Recently, MCNP® version 6.1 (MCNP6.1) has been released with extended cross-sections for low-energy electron/photon, subshell photoelectric cross-sections, and more detailed atomic relaxation data than the previous versions. With this new feature, the atomic relaxation process of MCNP6.1 has not been fully tested yet with its new physics library (eprdata12) that is based on the Evaluated Atomic Data Library (EADL). In this study, MCNP6.1 was compared with GATEv7.2, PENELOPE2014, and EGSnrc that have been often used to simulate low-energy atomic relaxation processes. The simulations were performed to acquire both photon and electron spectra produced by interactions of 15 keV electrons or photons with a 10-nm-thick gold nano-slab. The photon-induced fluorescence X-rays from MCNP6.1 fairly agreed with those from GATEv7.2 and PENELOPE2014, while the electron-induced fluorescence X-rays of the four codes showed more or less discrepancies. A coincidence was observed in the photon-induced Auger electrons simulated by MCNP6.1 and GATEv7.2. A recent release of MCNP6.1 with eprdata12 can be used to simulate the photon-induced atomic relaxation.

Purpose: The proton boron fusion reaction is a reaction that describes the creation of three alpha particles as the result of the interaction of a proton incident upon a 11B target. Theoretically, the proton boron fusion reaction is a desirable reaction for radiation therapy applications in that, with the appropriate boron delivery agent, it could potentially combine the localized dose delivery protons exhibit (Bragg peak) and the local deposition of high LET alpha particles in cancerous sites. Previous efforts have shown significant dose enhancement using the proton boron fusion reaction; the overarching purpose of this work is an attempt tomore » validate previous Monte Carlo results of the proton boron fusion reaction. Methods: The proton boron fusion reaction, 11B(p, 3α), is investigated using MCNP6 to assess the viability for potential use in radiation therapy. Simple simulations of a proton pencil beam incident upon both a water phantom and a water phantom with an axial region containing 100ppm boron were modeled using MCNP6 in order to determine the extent of the impact boron had upon the calculated energy deposition. Results: The maximum dose increase calculated was 0.026% for the incident 250 MeV proton beam scenario. The MCNP simulations performed demonstrated that the proton boron fusion reaction rate at clinically relevant boron concentrations was too small in order to have any measurable impact on the absorbed dose. Conclusion: For all MCNP6 simulations conducted, the increase of absorbed dose of a simple water phantom due to the 11B(p, 3α) reaction was found to be inconsequential. In addition, it was determined that there are no good evaluations of the 11B(p, 3α) reaction for use in MCNPX/6 and further work should be conducted in cross section evaluations in order to definitively evaluate the feasibility of the proton boron fusion reaction for use in radiation therapy applications.« less

The expanding clinical use of low-energy photon emitting 125I and 103Pd seeds in recent years has led to renewed interest in their dosimetric properties. Numerous papers pointed out that higher accuracy could be obtained in Monte Carlo simulations by utilizing newer libraries for the low-energy photon cross-sections, such as XCOM and EPDL97. The recently developed PENELOPE 2001 Monte Carlo code is user friendly and incorporates photon cross-section data from the EPDL97. The code has been verified for clinical dosimetry of high-energy electron and photon beams, but has not yet been tested at low energies. In the present work, we have benchmarked the PENELOPE code for 10-150 keV photons. We computed radial dose distributions from 0 to 10 cm in water at photon energies of 10-150 keV using both PENELOPE and MCNP4C with either DLC-146 or DLC-200 cross-section libraries, assuming a point source located at the centre of a 30 cm diameter and 20 cm length cylinder. Throughout the energy range of simulated photons (except for 10 keV), PENELOPE agreed within statistical uncertainties (at worst +/- 5%) with MCNP/DLC-146 in the entire region of 1-10 cm and with published EGS4 data up to 5 cm. The dose at 1 cm (or dose rate constant) of PENELOPE agreed with MCNP/DLC-146 and EGS4 data within approximately +/- 2% in the range of 20-150 keV, while MCNP/DLC-200 produced values up to 9% lower in the range of 20-100 keV than PENELOPE or the other codes. However, the differences among the four datasets became negligible above 100 keV.

The present work has simulated the photon and electron transport in a Theratron 780 (MDS Nordion) (60)Co radiotherapy unit, using the Monte Carlo transport code, MCNP (Monte Carlo N-Particle), version 5. In order to become computationally more efficient in view of taking part in the practical field of radiotherapy treatment planning, this work is focused mainly on the analysis of dose results and on the required computing time of different tallies applied in the model to speed up calculations.

A MCNP6 dosimetry model is presented for the Clinical Neutron Therapy System (CNTS) at the University of Washington. In the CNTS, fast neutrons are generated by a 50.5 MeV proton beam incident on a 10.5 mm thick Be target. The production, scattering and absorption of neutrons, photons, and other particles are explicitly tracked throughout the key components of the CNTS, including the target, primary collimator, flattening filter, monitor unit ionization chamber, and multi-leaf collimator. Simulations of the open field tissue maximum ratio (TMR), percentage depth dose profiles, and lateral dose profiles in a 40 cm × 40 cm × 40 cm water phantom are in good agreement with ionization chamber measurements. For a nominal 10 × 10 field, the measured and calculated TMR values for depths of 1.5 cm, 5 cm, 10 cm, and 20 cm (compared to the dose at 1.7 cm) are within 0.22%, 2.23%, 4.30%, and 6.27%, respectively. For the three field sizes studied, 2.8 cm × 2.8 cm, 10.4 cm × 10.3 cm, and 28.8 cm × 28.8 cm, a gamma test comparing the measured and simulated percent depth dose curves have pass rates of 96.4%, 100.0%, and 78.6% (depth from 1.5 to 15 cm), respectively, using a 3% or 3 mm agreement criterion. At a representative depth of 10 cm, simulated lateral dose profiles have in-field (⩾ 10% of central axis dose) pass rates of 89.7% (2.8 cm × 2.8 cm), 89.6% (10.4 cm × 10.3 cm), and 100.0% (28.8 cm × 28.8 cm) using a 3% and 3 mm criterion. The MCNP6 model of the CNTS meets the minimum requirements for use as a quality assurance tool for treatment planning and provides useful insights and information to aid in the advancement of fast neutron therapy.

The purpose of this project was to determine the fission multiplicity distribution, p(v), for the Beryllium Reflected Plutonium (BeRP) ball and to determine whether or not it changed appreciably for various High Density Polyethylene (HDPE) reflected configurations. The motivation for this project was to determine whether or not the average number of neutrons emitted per fission, v, changed significantly enough to reduce the discrepancy between MCNP6 and Robba, Dowdy, Atwater (RDA) point kinetic model estimates of multiplication. The energy spectrum of neutrons that induced fissions in the BeRP ball, NIF (E), was also computed in order to determine the averagemore » energy of neutrons inducing fissions, NIF . p(v) was computed using the FMULT card, NIF (E) and NIF were computed using an F4 tally with an FM tally modifier (F4/FM) card, and the multiplication factor, k eff, was computed using the KCODE card. Although NIF (E) changed significantly between bare and HDPE reflected configurations of the BeRP ball, the change in p(v), and thus the change in v, was insignificant. This is likely due to a difference between the way that NIF is computed using the FMULT and F4/FM cards. The F4/FM card indicated that NIF (E) was essentially Watt-fission distributed for a bare configuration and highly thermalized for all HDPE reflected configurations, while the FMULT card returned an average energy between 1 and 2 MeV for all configurations, which would indicate that the spectrum is Watt-fission distributed, regardless of the amount of HDPE reflector. The spectrum computed with the F4/FM cards is more physically meaningful and so the discrepancy between it and the FMULT card result is being investigated. It is hoped that resolving the discrepancy between the FMULT and F4/FM card estimates of NIF(E) will provide better v estimates that will lead to RDA multiplication estimates that are in better agreement with MCNP6 simulations.« less

Lost wax casting is an industrial process which permits the transmutation into metal of models made in wax. The wax model is covered with a silicaceous shell of the required thickness and once this shell is built the set is heated and wax melted. Liquid metal is then cast into the shell replacing the wax. When the metal is cool, the shell is broken away in order to recover the metallic piece. In this process zircon sands are used for the preparation of the silicaceous shell. These sands have varying concentrations of natural radionuclides: 238U, 232Th and 235U together with their progenics. The zircon sand is distributed in bags of 50 kg, and 30 bags are on a pallet, weighing 1,500 kg. The pallets with the bags have dimensions 80 cm x 120 cm x 80 cm, and constitute the radiation source in this case. The only pathway of exposure to workers in the store house is external radiation. In this case there is no dust because the bags are closed and covered by plastic, the store house has a good ventilation rate and so radon accumulation is not possible. The workers do not touch with their hands the bags and consequently skin contamination will not take place. In this study all situations of external irradiation to the workers have been considered; transportation of the pallets from vehicle to store house, lifting the pallets to the shelf, resting of the stock on the shelf, getting down the pallets, and carrying the pallets to production area. Using MCNP-4C exposure situations have been simulated, considering that the source has a homogeneous composition, the minimum stock in the store house is constituted by 7 pallets, and the several distances between pallets and workers when they are at work. The photons flux obtained by MCNP-4C is multiplied by the conversion factor of Flux to Kerma for air by conversion factor to Effective Dose by Kerma unit, and by the number of emitted photons. Those conversion factors are obtained of ICRP 74 table 1 and table 17 respectively. This

Purpose: To investigate the dose distributions in water cylinders simulating patients undergoing Interventional Radiological examinations. Method: The irradiation geometry consisted of an x-ray source, dose-area-product chamber, and image intensifier as currently used in Interventional Radiology. Water cylinders of diameters ranging between 17 and 30 cm were used to simulate patients weighing between 20 and 90 kg. X-ray spectra data with peak x-ray tube voltages ranging from 60 to 120 kV were generated using XCOMP3R. Radiation dose distributions inside the water cylinder (Dw) were obtained using MCNP5. The depth dose distribution along the x-ray beam central axis was normalized to free-in-air air kerma (AK) that is incident on the phantom. Scattered radiation within the water cylinders but outside the directly irradiated region was normalized to the dose at the edge of the radiation field. The total absorbed energy to the directly irradiated volume (Ep) and indirectly irradiated volume (Es) were also determined and investigated as a function of x-ray tube voltage and phantom size. Results: At 80 kV, the average Dw/AK near the x-ray entrance point was 1.3. The ratio of Dw near the entrance point to Dw near the exit point increased from ~ 26 for the 17 cm water cylinder to ~ 290 for the 30 cm water cylinder. At 80 kV, the relative dose for a 17 cm water cylinder fell to 0.1% at 49 cm away from the central ray of the x-ray beam. For a 30 cm water cylinder, the relative dose fell to 0.1% at 53 cm away from the central ray of the x-ray beam. At a fixed x-ray tube voltage of 80 kV, increasing the water cylinder diameter from 17 to 30 cm increased the Es/(Ep+Es) ratio by about 50%. At a fixed water cylinder diameter of 24 cm, increasing the tube voltage from 60 kV to 120 kV increased the Es/(Ep+Es) ratio by about 12%. The absorbed energy from scattered radiation was between 20-30% of the total energy absorbed by the water cylinder, and was affected more by patient size

MCNP has been enhanced to perform proton transport using a multigroup Fokker Planck (MGFP) algorithm with primary emphasis on proton radiography simulations. The new method solves the Fokker Planck approximation to the Boltzmann transport equation for the small angle multiple scattering portion of proton transport. Energy loss is accounted for by applying a group averaged stopping power over each transport step. Large angle scatter and non-inelastic events are treated as extinction. Comparisons with the more rigorous LAHET code show agreement to a few per cent for the total transmitted currents. The angular distributions through copper and low Z compounds showmore » good agreement between LAHET and MGFP with the MGFP method being slightly less forward peaked and without the large angle tails apparent in the LAHET simulation. Suitability of this method for proton radiography simulations is shown for a simple problem of a hole in a copper slab. LAHET and MGFP calculations of position, angle and energy through more complex objects are presented.« less

To ensure the reliability of simulation tools, verification and comparison should be made regularly. This paper describes the work performed in order to compare the neutron transport treatment in MCNP6.1 and GEANT4-10.3 in the thermal energy range. This work focuses on the thermal neutron scattering processes for several potential materials which would be involved in the neutron source designs of Compact Accelerator-based Neutrons Sources (CANS), such as beryllium metal, beryllium oxide, polyethylene, graphite, para-hydrogen, light water, heavy water, aluminium and iron. Both thermal scattering law and free gas model, coming from the evaluated data library ENDF/B-VII, were considered. It was observed that the GEANT4.10.03-patch2 version was not able to account properly the coherent elastic process occurring in crystal lattice. This bug is treated in this work and it should be included in the next release of the code. Cross section sampling and integral tests have been performed for both simulation codes showing a fair agreement between the two codes for most of the materials except for iron and aluminium.

The major components in the x-ray photon beam path of the treatment head of the VARIAN Clinac 2300 EX medical linear accelerator were modeled and simulated using the Monte Carlo N-Particle radiation transport computer code (MCNP). Simulated components include x-ray target, primary conical collimator, x-ray beam flattening filter and secondary collimators. X-ray photon energy spectra and angular distributions were calculated using the model. The x-ray beam emerging from the secondary collimators were scored by considering the total x-ray spectra from the target as the source of x-rays at the target position. The depth dose distribution and dose profiles at different depths and field sizes have been calculated at a nominal operating potential of 6 MV and found to be within acceptable limits. It is concluded that accurate specification of the component dimensions, composition and nominal accelerating potential gives a good assessment of the x-ray energy spectra.

In recent years voxel phantoms have been developed on the basis of tomographic data of real individuals allowing new sets of conversion coefficients to be calculated for effective dose. Progress in radiation studies brought ICRP to revise its recommendations and a new report, already circulated in draft form, is expected to change the actual effective dose evaluation method. In the present paper the voxel phantom NORMAN developed at HPA, formerly NRPB, was employed with MCNP Monte Carlo code. A modified version of the phantom, NORMAN-05, was developed to take into account the new set of tissues and weighting factors proposed in the cited ICRP draft. Air kerma to organ equivalent dose and effective dose conversion coefficients for antero-posterior and postero-anterior parallel photon beam irradiations, from 20 keV to 10 MeV, have been calculated and compared with data obtained in other laboratories using different numerical phantoms. Obtained results are in good agreement with published data with some differences for the effective dose calculated employing the proposed new tissue weighting factors set in comparison with previous evaluations based on the ICRP 60 report.

In this study, an approximate symmetric cylindrical manganese bath system with equal diameter and height was appraised using a Monte Carlo simulation. For nine sizes of the tank filled with MnSO(4).H(2)O solution of three different concentrations, the necessary correction factors involved in the absolute measurement of neutron emission rate were determined by a detailed modelling of the MCNP4C code with the ENDF/B-VII.0 neutron cross section data library. The results obtained were also used to determine the optimum dimensions of the bath for each concentration of solution in the calibration of (241)Am-Be and (252)Cf sources. Also, the amount of gamma radiation produced as a result of (n,γ) the reaction with the nuclei of the manganese sulphate solution that escaped from the boundary of each tank was evaluated. This gamma can be important for the background in NaI(Tl) detectors and issues concerned with radiation protection.

Purpose: Different versions of MCNP code are widely used for dosimetry purposes. The purpose of this study is to compare different versions of the MCNP codes in dosimetric evaluation of different brachytherapy sources. Methods: The TG-43 parameters such as dose rate constant, radial dose function, and anisotropy function of different brachytherapy sources, i.e. Pd-103, I-125, Ir-192, and Cs-137 were calculated in water phantom. The results obtained by three versions of Monte Carlo codes (MCNP4C, MCNPX, MCNP5) were compared for low and high energy brachytherapy sources. Then the cross section library of MCNP4C code was changed to ENDF/B-VI release 8 whichmore » is used in MCNP5 and MCNPX codes. Finally, the TG-43 parameters obtained using the MCNP4C-revised code, were compared with other codes. Results: The results of these investigations indicate that for high energy sources, the differences in TG-43 parameters between the codes are less than 1% for Ir-192 and less than 0.5% for Cs-137. However for low energy sources like I-125 and Pd-103, large discrepancies are observed in the g(r) values obtained by MCNP4C and the two other codes. The differences between g(r) values calculated using MCNP4C and MCNP5 at the distance of 6cm were found to be about 17% and 28% for I-125 and Pd-103 respectively. The results obtained with MCNP4C-revised and MCNPX were similar. However, the maximum difference between the results obtained with the MCNP5 and MCNP4C-revised codes was 2% at 6cm. Conclusion: The results indicate that using MCNP4C code for dosimetry of low energy brachytherapy sources can cause large errors in the results. Therefore it is recommended not to use this code for low energy sources, unless its cross section library is changed. Since the results obtained with MCNP4C-revised and MCNPX were similar, it is concluded that the difference between MCNP4C and MCNPX is their cross section libraries.« less

Many studies on biological effects of neutrons involve dose responses of neutrons, which rely on accurately determined absorbed doses in the irradiated cells or living organisms. Absorbed doses are difficult to measure, and are commonly surrogated with doses measured using separate detectors. The present work describes the determination of doses absorbed in the cell layer underneath a medium column (DA) and the doses absorbed in an ionization chamber (DE) from neutrons through computer simulations using the MCNP-5 code, and the subsequent determination of the conversion coefficients R (= DA/DE). It was found that R in general decreased with increase in the medium thickness, which was due to elastic and inelastic scattering. For 2-MeV neutrons, conspicuous bulges in R values were observed at medium thicknesses of about 500, 1500, 2500 and 4000 μm, and these were attributed to carbon, oxygen and nitrogen nuclei, and were reflections of spikes in neutron interaction cross sections with these nuclei. For 0.1-MeV neutrons, no conspicuous bulges in R were observed (except one at ~2000 μm that was due to photon interactions), which was explained by the absence of prominent spikes in the interaction cross-sections with these nuclei for neutron energies <0.1 MeV. The ratio R could be increased by ~50% for small medium thickness if the incident neutron energy was reduced from 2 MeV to 0.1 MeV. As such, the absorbed doses in cells (DA) would vary with the incident neutron energies, even when the absorbed doses shown on the detector were the same. PMID:27380801

Murine models are useful for targeted radiotherapy pre-clinical experiments. These models can help to assess the potential interest of new radiopharmaceuticals. In this study, we developed a voxel-based mouse for dosimetric estimates. A female nude mouse (30 g) was frozen and cut into slices. High-resolution digital photographs were taken directly on the frozen block after each section. Images were segmented manually. Monoenergetic photon or electron sources were simulated using the MCNP4c2 Monte Carlo code for each source organ, in order to give tables of S-factors (in Gy Bq-1 s-1) for all target organs. Results obtained from monoenergetic particles were then used to generate S-factors for several radionuclides of potential interest in targeted radiotherapy. Thirteen source and 25 target regions were considered in this study. For each source region, 16 photon and 16 electron energies were simulated. Absorbed fractions, specific absorbed fractions and S-factors were calculated for 16 radionuclides of interest for targeted radiotherapy. The results obtained generally agree well with data published previously. For electron energies ranging from 0.1 to 2.5 MeV, the self-absorbed fraction varies from 0.98 to 0.376 for the liver, and from 0.89 to 0.04 for the thyroid. Electrons cannot be considered as 'non-penetrating' radiation for energies above 0.5 MeV for mouse organs. This observation can be generalized to radionuclides: for example, the beta self-absorbed fraction for the thyroid was 0.616 for I-131; absorbed fractions for Y-90 for left kidney-to-left kidney and for left kidney-to-spleen were 0.486 and 0.058, respectively. Our voxel-based mouse allowed us to generate a dosimetric database for use in preclinical targeted radiotherapy experiments.

A set of conversion coefficients from kerma free-in-air to the organ-absorbed dose are presented for external monoenergetic photon beams from 10 keV to 10 MeV based on the Chinese mathematical phantom, a whole-body mathematical phantom model. The model was developed based on the methods of the Oak Ridge National Laboratory mathematical phantom series and data from the Chinese Reference Man and the Reference Asian Man. This work is carried out to obtain the conversion coefficients based on this model, which represents the characteristics of the Chinese population, as the anatomical parameters of the Chinese are different from those of Caucasians. Monte Carlo simulation with MCNP code is carried out to calculate the organ dose conversion coefficients. Before the calculation, the effects from the physics model and tally type are investigated, considering both the calculation efficiency and precision. In the calculation irradiation conditions include anterior-posterior, posterior-anterior, right lateral, left lateral, rotational and isotropic geometries. Conversion coefficients from this study are compared with those recommended in the Publication 74 of International Commission on Radiological Protection (ICRP74) since both the sets of data are calculated with mathematical phantoms. Overall, consistency between the two sets of data is observed and the difference for more than 60% of the data is below 10%. However, significant deviations are also found, mainly for the superficial organs (up to 65.9%) and bone surface (up to 66%). The big difference of the dose conversion coefficients for the superficial organs at high photon energy could be ascribed to kerma approximation for the data in ICRP74. Both anatomical variations between races and the calculation method contribute to the difference of the data for bone surface.

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

The MCNP Monte Carlo transport code is used by the Los Alamos National Laboratory Health and Safety Division for a broad spectrum of radiation shielding calculations. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with the Kansas State Univ. (KSU) photon skyshine experiment of 1977. The KSU experiment for the unshielded source geometry was simulated in great detail to include the contribution of groundshine, in-silo photon scatter, and the effect of spectral degradation in the source capsule. The standard deviation of the KSUmore » experimental data was stated to be 7%, while the statistical uncertainty of the simulation was kept at or under 1%. The results of the simulation agreed closely with the experimental data, generally to within 6%. At distances of under 100 m from the silo, the modeling of the in-silo scatter was crucial to achieving close agreement with the experiment. Specifically, scatter off the top layer of the source cask accounted for [approximately]12% of the dose at 50 m. At distance >300m, using the [sup 60]Co line spectrum led to a dose overresponse as great as 19% at 700 m. It was necessary to use the actual source spectrum, which includes a Compton tail from photon collisions in the source capsule, to achieve close agreement with experimental data. These results highlight the importance of using Monte Carlo transport techniques to account for the nonideal features of even simple experiments''.« less

In this work, the general purpose Monte Carlo N-particle radiation transport computer code (MCNP-4C) was used for the simulation of X-ray spectra in diagnostic radiology. The electron's path in the target was followed until its energy was reduced to 10 keV. A user-friendly interface named 'diagnostic X-ray spectra by Monte Carlo simulation (DXRaySMCS)' was developed to facilitate the application of MCNP-4C code for diagnostic radiology spectrum prediction. The program provides a user-friendly interface for: (i) modifying the MCNP input file, (ii) launching the MCNP program to simulate electron and photon transport and (iii) processing the MCNP output file to yield a summary of the results (relative photon number per energy bin). In this article, the development and characteristics of DXRaySMCS are outlined. As part of the validation process, output spectra for 46 diagnostic radiology system settings produced by DXRaySMCS were compared with the corresponding IPEM78. Generally, there is a good agreement between the two sets of spectra. No statistically significant differences have been observed between IPEM78 reported spectra and the simulated spectra generated in this study.

Cosmic neutron and photon fluxes are known to scale exponentially with elevation. Consequently, cosmic neutron elevation scaling was implemented for use with the background-source option shortly after its introduction into MCNP6, whereby the neutron flux weight factor was adjusted by the elevation scaling factor when the user-specified elevation differed from the selected background.dat grid-point elevation. At the same time, an elevation scaling factor was suggested for the cosmic photon flux, however, cosmic photon elevation scaling is complicated by the fact that the photon background consists of two components: cosmic and terrestrial. Previous versions of the background.dat file did not provide any way to separate these components. With Rel. 4 of this file in 2015, two new columns were added that provide the energy grid and differential cosmic photon flux separately from the total photon flux. Here we show that the cosmic photon flux component can now be scaled independently and combined with the terrestrial component to form the total photon flux at a user-specified elevation in MCNP6. Cosmic background fluxes also scale with the solar cycle due to solar modulation. This modulation has been shown to be nearly sinusoidal over time, with an inverse effect - increased modulation leads to a decrease in cosmic fluxes. This effect was initially included with the cosmic source option in MCNP6 and has now been extended for use with the background source option when: (1) the date is specified in the background.dat file, and (2) when the user specifies a date on the source definition card. A description of the cosmic-neutron/photon date scaling feature will be presented along with scaling results for past and future date extrapolations.

The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combinations to investigate the effect of tube voltage, target material and filter thickness on x-ray spectra in the diagnostic radiology and mammography energy ranges. The simulated x-ray spectra were compared with experimental measurements and spectra calculated by IPEM report number 78. In addition, the anode heel effect and off-axis x-ray spectra were assessed for different anode angles and target materials and the results were compared with EGS4-based Monte Carlo simulations and measured data. Quantitative evaluation of the differences between our Monte Carlo simulated and comparison spectra was performed using student's t-test statistical analysis. Generally, there is a good agreement between the simulated x-ray and comparison spectra, although there are systematic differences between the simulated and reference spectra especially in the K-characteristic x-rays intensity. Nevertheless, no statistically significant differences have been observed between IPEM spectra and the simulated spectra. It has been shown that the difference between MCNP simulated spectra and IPEM spectra in the low energy range is the result of the overestimation of characteristic photons following the normalization procedure. The transmission curves produced by MCNP4C have good agreement with the IPEM report especially for tube voltages of 50 kV and 80 kV. The systematic discrepancy for higher tube voltages is the result of systematic differences between the corresponding spectra.

A cyclic electron accelerator Microtron MT-25 will be installed in Havana, Cuba. Electrons, neutrons and gamma radiation up to 25 MeV can be produced in the MT-25. A detailed shielding analysis for the bunker is carried out using two ways: the NCRP-51 Report and the Monte Carlo Method (MCNP-4C Code). The walls and ceiling thicknesses are estimated with dose constraints of 0.5 and 20 mSv y(-1), respectively, and an area occupancy factor of 1/16. Both results are compared and a preliminary bunker design is shown. Copyright 2004 Oxford University Press

The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.

In this study, element analysis and the mass attenuation coefficient for matrixes of gold, bronze and water with various impurities and the concentrations of heavy metals (Cu, Mn, Pb and Zn) are evaluated and calculated by the MCNP simulation code for photons emitted from Barium-133, Americium-241 and sources with energies between 1 and 100 keV. The MCNP data are compared with the experimental data and WinXCom code simulated results by Medhat. The results showed that the obtained results of bronze and gold matrix are in good agreement with the other methods for energies above 40 and 60 keV, respectively. However for water matrixes with various impurities, there is a good agreement between the three methods MCNP, WinXCom and the experimental one in low and high energies.

Improved cross-sections in a new version of the Monte-Carlo N-particle (MCNP) code may eliminate discrepancies between radial dose functions (as defined by American Association of Physicists in Medicine Task Group 43) derived from Monte-Carlo simulations of low-energy photon-emitting brachytherapy sources and those from measurements on the same sources with thermoluminescent dosimeters. This is demonstrated for two 125I brachytherapy seed models, the Implant Sciences Model ISC3500 (I-Plant) and the Amersham Health Model 6711, by simulating their radial dose functions with two versions of MCNP, 4c2 and 5.

Here, cosmic neutron and photon fluxes are known to scale exponentially with elevation. Consequently, cosmic neutron elevation scaling was implemented for use with the background-source option shortly after its introduction into MCNP6, whereby the neutron flux weight factor was adjusted by the elevation scaling factor when the user-specified elevation differed from the selected background.dat grid-point elevation. At the same time, an elevation scaling factor was suggested for the cosmic photon flux, however, cosmic photon elevation scaling is complicated by the fact that the photon background consists of two components: cosmic and terrestrial. Previous versions of the background.dat file did notmore » provide any way to separate these components. With Rel. 4 of this file in 2015, two new columns were added that provide the energy grid and differential cosmic photon flux separately from the total photon flux. Here we show that the cosmic photon flux component can now be scaled independently and combined with the terrestrial component to form the total photon flux at a user-specified elevation in MCNP6.« less

A novel object-oriented modular mapping methodology for externally coupled neutronics–thermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. In addition, it performs the basis mapping from themore » combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers.« less

Here, cosmic neutron and photon fluxes are known to scale exponentially with elevation. Consequently, cosmic neutron elevation scaling was implemented for use with the background-source option shortly after its introduction into MCNP6, whereby the neutron flux weight factor was adjusted by the elevation scaling factor when the user-specified elevation differed from the selected background.dat grid-point elevation. At the same time, an elevation scaling factor was suggested for the cosmic photon flux, however, cosmic photon elevation scaling is complicated by the fact that the photon background consists of two components: cosmic and terrestrial. Previous versions of the background.dat file did notmore » provide any way to separate these components. With Rel. 4 of this file in 2015, two new columns were added that provide the energy grid and differential cosmic photon flux separately from the total photon flux. Here we show that the cosmic photon flux component can now be scaled independently and combined with the terrestrial component to form the total photon flux at a user-specified elevation in MCNP6.« less

The Monte Carlo computer codes EGS4 and MCNP were used to develop a theoretical model of a 180 degrees geometry in vivo X-ray fluorescence system for the measurement of platinum concentration in head and neck tumors. The model included specification of the photon source, collimators, phantoms and detector. Theoretical results were compared and evaluated against X-ray fluorescence data obtained experimentally from an existing system developed by the Swansea In Vivo Analysis and Cancer Research Group. The EGS4 results agreed well with the MCNP results. However, agreement between the measured spectral shape obtained using the experimental X-ray fluorescence system and the simulated spectral shape obtained using the two Monte Carlo codes was relatively poor. The main reason for the disagreement between the results arises from the basic assumptions which the two codes used in their calculations. Both codes assume a "free" electron model for Compton interactions. This assumption will underestimate the results and invalidates any predicted and experimental spectra when compared with each other.

A set of conversion coefficients from kerma free-in-air to the organ absorbed dose for external photon beams from 10 keV to 10 MeV are presented based on a newly developed voxel mouse model, for the purpose of radiation effect evaluation. The voxel mouse model was developed from colour images of successive cryosections of a normal nude male mouse, in which 14 organs or tissues were segmented manually and filled with different colours, while each colour was tagged by a specific ID number for implementation of mouse model in Monte Carlo N-particle code (MCNP). Monte Carlo simulation with MCNP was carried out to obtain organ dose conversion coefficients for 22 external monoenergetic photon beams between 10 keV and 10 MeV under five different irradiation geometries conditions (left lateral, right lateral, dorsal-ventral, ventral-dorsal, and isotropic). Organ dose conversion coefficients were presented in tables and compared with the published data based on a rat model to investigate the effect of body size and weight on the organ dose. The calculated and comparison results show that the organ dose conversion coefficients varying the photon energy exhibits similar trend for most organs except for the bone and skin, and the organ dose is sensitive to body size and weight at a photon energy approximately <0.1 MeV.

During heavy ion irradiation therapy the patient has to be located exactly at the right position to make sure that the Bragg peak occurs in the tumour. The patient has to be moved in the range of millimetres to scan the ill tissue. For that reason a special table was developed which allows exact positioning. The electronic control can be located outside the surgery. But that has some disadvantage for the construction. To keep the system compact it would be much more comfortable to put the electronic control inside the surgery. As a lot of high energetic secondary particles are produced during the therapy causing a high dose in the room it is important to find positions with low dose rates. Therefore, investigations are needed where the electronic devices should be located to obtain a minimum of radiation, help to prevent the failure of sensitive devices. The dose rate was calculated for carbon ions with different initial energy and protons over the entire therapy room with Monte Carlo particle tracking using MCNP6. The types of secondary particles were identified and the dose rate for a thin silicon layer and an electronic mixture material was determined. In addition, the shielding effect of several selected material layers was calculated using MCNP6.

Neutron measurements using neutron activation systems are an essential part of the diagnostic system at large fusion machines such as JET and ITER. Nuclear data is used to infer the neutron yield. Consequently, high-quality nuclear data is essential for the proper determination of the neutron yield and fusion power. However, uncertainties due to nuclear data are not fully taken into account in uncertainty analysis for neutron yield calibrations using activation foils. This paper investigates the neutron yield uncertainty due to nuclear data using the so-called Total Monte Carlo Method. The work is performed using a detailed MCNP model of the JET fusion machine; the uncertainties due to the cross-sections and angular distributions in JET structural materials, as well as the activation cross-sections in the activation foils, are analysed. It is found that a significant contribution to the neutron yield uncertainty can come from uncertainties in the nuclear data.

In this study, the multi-threading performance of the Geant4, MCNP6, and PHITS codes was evaluated as a function of the number of threads (N) and the complexity of the tetrahedral-mesh phantom. For this, three tetrahedral-mesh phantoms of varying complexity (simple, moderately complex, and highly complex) were prepared and implemented in the three different Monte Carlo codes, in photon and neutron transport simulations. Subsequently, for each case, the initialization time, calculation time, and memory usage were measured as a function of the number of threads used in the simulation. It was found that for all codes, the initialization time significantly increased with the complexity of the phantom, but not with the number of threads. Geant4 exhibited much longer initialization time than the other codes, especially for the complex phantom (MRCP). The improvement of computation speed due to the use of a multi-threaded code was calculated as the speed-up factor, the ratio of the computation speed on a multi-threaded code to the computation speed on a single-threaded code. Geant4 showed the best multi-threading performance among the codes considered in this study, with the speed-up factor almost linearly increasing with the number of threads, reaching ~30 when N = 40. PHITS and MCNP6 showed a much smaller increase of the speed-up factor with the number of threads. For PHITS, the speed-up factors were low when N = 40. For MCNP6, the increase of the speed-up factors was better, but they were still less than ~10 when N = 40. As for memory usage, Geant4 was found to use more memory than the other codes. In addition, compared to that of the other codes, the memory usage of Geant4 more rapidly increased with the number of threads, reaching as high as ~74 GB when N = 40 for the complex phantom (MRCP). It is notable that compared to that of the other codes, the memory usage of PHITS was much lower, regardless of both the complexity of the

Permanent implantation of low energy (20-40 keV) photon emitting radioactive seeds to treat prostate cancer is an important treatment option for patients. In order to produce accurate implant brachytherapy treatment plans, the dosimetry of a single source must be well characterized. Monte Carlo based transport calculations can be used for source characterization, but must have up to date cross section libraries to produce accurate dosimetry results. This work benchmarks the MCNP code and its photon cross section library for low energy photon brachytherapy applications. In particular, we calculate the emitted photon spectrum, air kerma, depth dose in water, and radial dose function for both 125I and 103Pd based seeds and compare to other published results. Our results show that MCNP's cross section library differs from recent data primarily in the photoelectric cross section for low energies and low atomic number materials. In water, differences as large as 10% in the photoelectric cross section and 6% in the total cross section occur at 125I and 103Pd photon energies. This leads to differences in the dose rate constant of 3% and 5%, and differences as large as 18% and 20% in the radial dose function for the 125I and 103Pd based seeds, respectively. Using a partially updated photon library, calculations of the dose rate constant and radial dose function agree with other published results. Further, the use of the updated photon library allows us to verify air kerma and depth dose in water calculations performed using MCNP's perturbation feature to simulate updated cross sections. We conclude that in order to most effectively use MCNP for low energy photon brachytherapy applications, we must update its cross section library. Following this update, the MCNP code system will be a very effective tool for low energy photon brachytherapy dosimetry applications.

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI formore » performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)« less

This work describes a simulation study investigating the application of neutron stimulated emission computed tomography (NSECT) for noninvasive 3D imaging of renal cancer in vivo. Using MCNP5 simulations, we describe a method of diagnosing renal cancer in the body by mapping the 3D distribution of elements present in tumors using the NSECT technique. A human phantom containing the kidneys and other major organs was modeled in MCNP5. The element composition of each organ was based on values reported in literature. The two kidneys were modeled to contain elements reported in renal cell carcinoma (RCC) and healthy kidney tissue. Simulated NSECT scans were executed to determine the 3D element distribution of the phantom body. Elements specific to RCC and healthy kidney tissue were then analyzed to identify the locations of the diseased and healthy kidneys and generate tomographic images of the tumor. The extent of the RCC lesion inside the kidney was determined using 3D volume rendering. A similar procedure was used to generate images of each individual organ in the body. Six isotopes were studied in this work - (32)S, (12)C, (23)Na, (14)N, (31)P and (39)K. The results demonstrated that through a single NSECT scan performed in vivo, it is possible to identify the location of the kidneys and other organs within the body, determine the extent of the tumor within the organ, and to quantify the differences between cancer and healthy tissue-related isotopes with p ≤ 0.05. All of the images demonstrated appropriate concentration changes between the organs, with some discrepancy observed in (31)P, (39)K and (23)Na. The discrepancies were likely due to the low concentration of the elements in the tissue that were below the current detection sensitivity of the NSECT technique.

Nuclear Criticality Safety (NCS) has performed calculations evaluating the effect of different interstitial materials on 5.0-kg of plutonium metal. As with all non-fissionable interstitials, the results here illustrate that it requires significant quantities of oil to be intimately mixed with plutonium, reflected by a thick layer of full-density water, to achieve the same reactivity as that of solid plutonium metal.

The United Kingdom's National Nuclear Laboratory (NNL) has developed a remote, non-electrical, radiation-mapping device known as RadBall(™), which can locate and quantify radioactive hazards within contaminated areas of the nuclear industry. RadBall(™) consists of a colander-like outer shell that houses a radiation-sensitive polymer sphere. The outer shell works to collimate radiation sources and those areas of the polymer sphere that are exposed react, becoming increasingly more opaque, in proportion to the absorbed dose. The polymer sphere is imaged in an optical-CT scanner, which produces a high resolution 3D map of optical attenuation coefficients. Subsequent analysis of the optical attenuation matrix provides information on the spatial distribution of sources in a given area forming a 3D characterization of the area of interest. RadBall(™) has no power requirements and can be positioned in tight or hard-to reach locations. The RadBall(™) technology has been deployed in a number of technology trials in nuclear waste reprocessing plants at Sellafield in the United Kingdom and facilities of the Savannah River National Laboratory (SRNL). This study focuses on the RadBall(™) testing and modeling accomplished at SRNL.

The United Kingdom’s National Nuclear Laboratory (NNL) has developed a remote, non-electrical, radiation-mapping device known as RadBall™, which can locate and quantify radioactive hazards within contaminated areas of the nuclear industry. RadBall™ consists of a colander-like outer shell that houses a radiation-sensitive polymer sphere. The outer shell works to collimate radiation sources and those areas of the polymer sphere that are exposed react, becoming increasingly more opaque, in proportion to the absorbed dose. The polymer sphere is imaged in an optical-CT scanner, which produces a high resolution 3D map of optical attenuation coefficients. Subsequent analysis of the optical attenuation matrix provides information on the spatial distribution of sources in a given area forming a 3D characterization of the area of interest. RadBall™ has no power requirements and can be positioned in tight or hard-to reach locations. The RadBall™ technology has been deployed in a number of technology trials in nuclear waste reprocessing plants at Sellafield in the United Kingdom and facilities of the Savannah River National Laboratory (SRNL). This study focuses on the RadBall™ testing and modeling accomplished at SRNL. PMID:21617740

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for gadolinium neutron capture therapy (GdNCT) related modeling. The validity of COG NCT model has been established for this model, and here the calculation was extended to analyze the effect of various gadolinium concentrations on dose distribution and cell-kill effect of the GdNCT modality and to determine the optimum therapeutic conditions for treating brain cancers. The computational results were compared with the widely used MCNP code. The differences between the COG and MCNP predictions were generally small and suggest that the COG code can be applied to similar research problems in NCT. Results for this study also showed that a concentration of 100 ppm gadolinium in the tumor was most beneficial when using an epithermal neutron beam.

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials withmore » different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)« less

Unwanted neutrons in radiation therapy treatments are typically generated by photonuclear reactions. High-energy beams emitted by medical Linear Accelerators (LinAcs) interact with high atomic number materials situated in the accelerator head and release neutrons. Since neutrons have a high relative biological effectiveness, even low neutron doses may imply significant exposure of patients. It is also important to study radioactivity induced by these photoneutrons when interacting with the different materials and components of the treatment head facility and the shielding room walls, since persons not present during irradiation (e.g. medical staff) may be exposed to them even when the accelerator is not operating. These problems are studied in this work in order to contribute to challenge the radiation protection in these treatment locations. The work has been performed by simulation using the latest state of the art of Monte-Carlo computer code MCNP6. To that, a detailed model of particles transport inside the bunker and treatment head has been carried out using a meshed geometry model. The LinAc studied is an Elekta Precise accelerator with a treatment photon energy of 15 MeV used at the Hospital Clinic Universitari de Valencia, Spain.

The development of a computational model for the treatment head of a medical electron accelerator (Elekta/Philips SL-18) by the Monte Carlo code mcnp-4C2 is discussed. The model includes the major components of the accelerator head and a pmma phantom representing the patient body. Calculations were performed for a 14 MeV electron beam impinging on the accelerator target and a 10 cmx10 cm beam area at the isocentre. The model was used in order to predict the neutron ambient dose equivalent at the isocentre level and moreover the neutron absorbed dose distribution within the phantom. Calculations were validated against experimental measurements performed by gold foil activation detectors. The results of this study indicated that the equivalent dose at tissues or organs adjacent to the treatment field due to photoneutrons could be up to 10% of the total peripheral dose, for the specific accelerator characteristics examined. Therefore, photoneutrons should be taken into account when accurate dose calculations are required to sensitive tissues that are adjacent to the therapeutic X-ray beam. The method described can be extended to other accelerators and collimation configurations as well, upon specification of treatment head component dimensions, composition and nominal accelerating potential.

The use of a hydrogen internal standard to enable the estimation of absolute mass during measurement of total body nitrogen by in vivo neutron activation is an established technique. Central to the technique is a determination of the H prompt gamma ray counts arising from the subject. In practice, interference counts from other sources--e.g., neutron shielding--are included. This study reports use of the Monte Carlo computer code, MCNP-4A, to investigate the interference counts arising from shielding both with and without a phantom containing a urea solution. Over a range of phantom size (depth 5 to 30 cm, width 20 to 40 cm), the counts arising from shielding increased by between 4% and 32% compared with the counts without a phantom. For any given depth, the counts increased approximately linearly with width. For any given width, there was little increase for depths exceeding 15 centimeters. The shielding counts comprised between 15% and 26% of those arising from the urea phantom. These results, although specific to the Swansea apparatus, suggest that extraneous hydrogen counts can be considerable and depend strongly on the subject's size.

The dose parameters for the beta-particle emitting 90Sr/90Y source for intravascular brachytherapy (IVBT) have been calculated by different investigators. At a distant distance from the source, noticeable differences are seen in these parameters calculated using different Monte Carlo codes. The purpose of this work is to quantify as well as to understand these differences. We have compared a series of calculations using an EGS4, an EGSnrc, and the MCNP Monte Carlo codes. Data calculated and compared include the depth dose curve for a broad parallel beam of electrons, and radial dose distributions for point electron sources (monoenergetic or polyenergetic) and for a real 90Sr/90Y source. For the 90Sr/90Y source, the doses at the reference position (2 mm radial distance) calculated by the three code agree within 2%. However, the differences between the dose calculated by the three codes can be over 20% in the radial distance range interested in IVBT. The difference increases with radial distance from source, and reaches 30% at the tail of dose curve. These differences may be partially attributed to the different multiple scattering theories and Monte Carlo models for electron transport adopted in these three codes. Doses calculated by the EGSnrc code are more accurate than those by the EGS4. The two calculations agree within 5% for radial distance <6 mm.

High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.

To evaluate conventional munition, we simulated an x-ray computed tomography (CT) system for generating radiographs from nominal x-ray energies of 6 or 9 megaelectron volts (MeV). CT simulations, informed by measured data, allow for optimization of both system design and acquisition techniques necessary to enhance image quality. MCNP6 radiographic simulation tools were used to model ideal detector responses (DR) that assume either (1) a detector response proportional to photon flux (N) or (2) a detector response proportional to energy flux (E). As scatter may become significant with MeV x-ray systems, simulations were performed with and without the inclusion of object scatter. Simulations were compared against measurements of a cylindrical munition component principally composed of HMX, tungsten and aluminum encased in carbon fiber. Simulations and measurements used a 6 MeV peak energy x-ray spectrum filtered with 3.175 mm of tantalum. A detector response proportional to energy which includes object scatter agrees to within 0.6 % of the measured line integral of the linear attenuation coefficient. Exclusion of scatter increases the difference between measurement and simulation to 5 %. A detector response proportional to photon flux agrees to within 20 % when object scatter is included in the simulation and 27 % when object scatter is excluded.

This work has been performed within the frame of the European Union ORAMED project (Optimisation of RAdiation protection for MEDical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

Molecular imaging technologies provide unique abilities to localise signs of disease before symptoms appear, assist in drug testing, optimize and personalize therapy, and assess the efficacy of treatment regimes for different types of cancer. Monte Carlo simulation packages are used as an important tool for the optimal design of detector systems. In addition they have demonstrated potential to improve image quality and acquisition protocols. Many general purpose (MCNP, Geant4, etc) or dedicated codes (SimSET etc) have been developed aiming to provide accurate and fast results. Special emphasis will be given to GATE toolkit. The GATE code currently under development by the OpenGATE collaboration is the most accurate and promising code for performing realistic simulations. The purpose of this article is to introduce the non expert reader to the current status of MC simulations in nuclear medicine and briefly provide examples of current simulated systems, and present future challenges that include simulation of clinical studies and dosimetry applications.

Theoretical calculations indicate that the value of the Feynman variance, Y2F for the emitted distribution of neutrons from ssionable exhibits a strong monotonic de- pendence on a the multiplication, M, of a quantity of special nuclear material. In 2012 we performed a series of measurements at the Passport Inc. facility using a 9- MeV bremsstrahlung CW beam of photons incident on small quantities of uranium with liquid scintillator detectors. For the set of objects studies we observed deviations in the expected monotonic dependence, and these deviations were later con rmed by MCNP simulations. In this report, we modify the theorymore » to account for the contri- bution from the initial photo- ssion and benchmark the new theory with a series of MCNP simulations on DU, LEU, and HEU objects spanning a wide range of masses and multiplication values.« less

At thermal energies, the scattering of neutrons in a system is complicated by the comparable velocities of the neutron and target, resulting in competing upscattering and downscattering events. The neutron wavelength is also similar in size to the target's interatomic spacing making the scattering process a quantum mechanical problem. Because of the complicated nature of scattering at low energies, the thermal data files in ACE format used in continuous-energy Monte Carlo codes are quite large { on the order of megabytes for a single temperature and material. In this paper, a new storage and sampling method is introduced that ismore » orders of magnitude less in size and is used to sample scattering parameters at any temperature on-the-fly. In addition to the reduction in storage, the need to pre-generate thermal scattering data tables at fine temperatures has been eliminated. This is advantageous for multiphysics simulations which may involve temperatures not known in advance. A new module was written for MCNP6 that bypasses the current S(α,β) table lookup in favor of the new format. The new on-the-fly sampling method was tested for graphite for two benchmark problems at ten temperatures: 1) an eigenvalue test with a fuel compact of uranium oxycarbide fuel homogenized into a graphite matrix, 2) a surface current test with a \\broomstick" problem with a monoenergetic point source. The largest eigenvalue difference was 152pcm for T= 1200K. For the temperatures and incident energies chosen for the broomstick problem, the secondary neutron spectrum showed good agreement with the traditional S(α,β) sampling method. These preliminary results show that sampling thermal scattering data on-the-fly is a viable option to eliminate both the storage burden of keeping thermal data at discrete temperatures and the need to know temperatures before simulation runtime.« less

This paper evaluates the accuracy of the single-event (SE) and condensed-history (CH) models of electron transport in MCNP6.1 when simulating characteristic Kα, total K (=Kα + Kβ) and Lα X-ray emission from thick targets bombarded by electrons with energies from 5 keV to 30 keV. It is shown that the MCNP6.1 implementation of the CH model for the K-shell impact ionization leads to underestimation of the K yield by 40% or more for the elements with atomic numbers Z < 15 and overestimation of the Kα yield by more than 40% for the elements with Z > 25. The Lα yields are underestimated by more than an order of magnitude in CH mode, because MCNP6.1 neglects X-ray emission caused by electron-impact ionization of L, M and higher shells in CH mode (the Lα yields calculated in CH mode reflect only X-ray fluorescence, which is mainly caused by photoelectric absorption of bremsstrahlung photons). The X-ray yields calculated by MCNP6.1 in SE mode (using ENDF/B-VII.1 library data) are more accurate: the differences of the calculated and experimental K yields are within the experimental uncertainties for the elements C, Al and Si, and the calculated Kα yields are typically underestimated by (20-30)% for the elements with Z > 25, whereas the Lα yields are underestimated by (60-70)% for the elements with Z > 49. It is also shown that agreement of the experimental X-ray yields with those calculated in SE mode is additionally improved by replacing the ENDF/B inner-shell electron-impact ionization cross sections with the set of cross sections obtained from the distorted-wave Born approximation (DWBA), which are also used in the PENELOPE code system. The latter replacement causes a decrease of the average relative difference of the experimental X-ray yields and the simulation results obtained in SE mode to approximately 10%, which is similar to accuracy achieved with PENELOPE. This confirms that the DWBA inner-shell impact ionization cross sections are significantly more

To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient model to accurately predict particle relative biological effectiveness (RBE) in cell cultures and in vivo. To illustrate the approach and highlight the impact of the larger scale physical environment (e.g. establishing charged particle equilibrium), we examined the RBE for DSB induction (RBEDSB) of x-rays, (137)Cs γ-rays, neutrons and light ions relative to γ-rays from (60)Co in monolayer cell cultures at various depths in water. Under normoxic conditions, we found that (137)Cs γ-rays are about 1.7% more effective at creating DSB than γ-rays from (60)Co (RBEDSB = 1.017) whereas 60-250 kV x-rays are 1.1 to 1.25 times more efficient at creating DSB than (60)Co. Under anoxic conditions, kV x-rays may have an RBEDSB up to 1.51 times as large as (60)Co γ-rays. Fission neutrons passing through monolayer cell cultures have an RBEDSB that ranges from 2.6 to 3.0 in normoxic cells, but may be as large as 9.93 for anoxic cells. For proton pencil beams, Monte Carlo simulations suggest an RBEDSB of about 1.2 at the tip of the Bragg peak and up to 1.6 a few mm beyond the Bragg peak. Bragg peak RBEDSB increases with decreasing oxygen concentration, which may create opportunities to apply proton dose painting to help address tumor hypoxia. Modeling of the particle RBE for DSB induction across multiple physical and biological scales has the potential to aid in the interpretation of laboratory experiments and provide useful information to advance the safety and effectiveness of hadron therapy in the treatment of cancer.

Here, a proper treatment of thermal neutron scattering requires accounting for chemical binding through a scattering law S(α,β,T). Monte Carlo codes sample the secondary neutron energy and angle after a thermal scattering event from probability tables generated from S(α,β,T) tables at discrete temperatures, requiring a large amount of data for multiscale and multiphysics problems with detailed temperature gradients. We have previously developed a method to handle this temperature dependence on-the-fly during the Monte Carlo random walk using polynomial expansions in 1/T to directly sample the secondary energy and angle. In this paper, the on-the-fly method is implemented into MCNP6 andmore » tested in both graphite-moderated and light water-moderated systems. The on-the-fly method is compared with the thermal ACE libraries that come standard with MCNP6, yielding good agreement with integral reactor quantities like k-eigenvalue and differential quantities like single-scatter secondary energy and angle distributions. The simulation runtimes are comparable between the two methods (on the order of 5–15% difference for the problems tested) and the on-the-fly fit coefficients only require 5–15 MB of total data storage.« less

A new method for analysis of uranium and radium in soils by gamma spectroscopy has been developed using VRF ("Visual RobFit") which, unlike traditional peak-search techniques, fits full-spectrum nuclide shapes with non-linear least-squares minimization of the chi-squared statistic. Gamma efficiency curves were developed for a 500 mL Marinelli beaker geometry as a function of soil density using MCNP. Collected spectra were then analyzed using the MCNP-generated efficiency curves and VRF to deconvolute the 90 keV peak complex of uranium and obtain 238U and 235U activities. 226Ra activity was determined either from the radon daughters if the equilibrium status is known, or directly from the deconvoluted 186 keV line. 228Ra values were determined from the 228Ac daughter activity. The method was validated by analysis of radium, thorium and uranium soil standards and by inter-comparison with other methods for radium in soils. The method allows for a rapid determination of whether a sample has been impacted by a man-made activity by comparison of the uranium and radium concentrations to those that would be expected from a natural equilibrium state.

This paper aims at comparing dosimetric assessments performed with three Monte Carlo codes: EGS4, MCNP4c2 and MCNPX2.5e, using a realistic voxel phantom, namely the Zubal phantom, in two configurations of exposure. The first one deals with an external irradiation corresponding to the example of a radiological accident. The results are obtained using the EGS4 and the MCNP4c2 codes and expressed in terms of the mean absorbed dose (in Gy per source particle) for brain, lungs, liver and spleen. The second one deals with an internal exposure corresponding to the treatment of a medullary thyroid cancer by 131I-labelled radiopharmaceutical. The results are obtained by EGS4 and MCNPX2.5e and compared in terms of S-values (expressed in mGy per kBq and per hour) for liver, kidney, whole body and thyroid. The results of these two studies are presented and differences between the codes are analysed and discussed.

This paper contains two parts revolving around Monte Carlo transport simulation on Intel Many Integrated Core coprocessors (MIC, also known as Xeon Phi). (1) MCNP 6.1 was recompiled into multithreading (OpenMP) and multiprocessing (MPI) forms respectively without modification to the source code. The new codes were tested on a 60-core 5110P MIC. The test case was FS7ONNi, a radiation shielding problem used in MCNP's verification and validation suite. It was observed that both codes became slower on the MIC than on a 6-core X5650 CPU, by a factor of 4 for the MPI code and, abnormally, 20 for the OpenMP code, and both exhibited limited capability of strong scaling. (2) We have recently added a Constructive Solid Geometry (CSG) module to our ARCHER code to provide better support for geometry modelling in radiation shielding simulation. The functions of this module are frequently called in the particle random walk process. To identify the performance bottleneck we developed a CSG proxy application and profiled the code using the geometry data from FS7ONNi. The profiling data showed that the code was primarily memory latency bound on the MIC. This study suggests that despite low initial porting e_ort, Monte Carlo codes do not naturally lend themselves to the MIC platform — just like to the GPUs, and that the memory latency problem needs to be addressed in order to achieve decent performance gain.

In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types—based on BGO, NaI and LaBr3 crystals is presented.

MCNP simulations have been run to evaluate the feasibility of using a combination of fast and thermal neutrons as a nondestructive method to measure of the compaction of the perlite insulation in the liquid hydrogen and oxygen cryogenic storage tanks at John F. Kennedy Space Center (KSC). Perlite is a feldspathic volcanic rock made up of the major elements Si, AI, Na, K and 0 along with some water. When heated it expands from four to twenty times its original volume which makes it very useful for thermal insulation. The cryogenic tanks at Kennedy Space Center are spherical with outer diameters of 69-70 feet and lined with a layer of expanded perlite with thicknesses on the order of 120 cm. There is evidence that some of the perlite has compacted over time since the tanks were built 1965, affecting the thermal properties and possibly also the structural integrity of the tanks. With commercially available portable neutron generators it is possible to produce simultaneously fluxes of neutrons in two energy ranges: fast (14 Me V) and thermal (25 me V). The two energy ranges produce complementary information. Fast neutrons produce gamma rays by inelastic scattering, which is sensitive to Fe and O. Thermal neutrons produce gamma rays by prompt gamma neutron activation (PGNA) and this is sensitive to Si, Al, Na, K and H. The compaction of the perlite can be measured by the change in gamma ray signal strength which is proportional to the atomic number densities of the constituent elements. The MCNP simulations were made to determine the magnitude of this change. The tank wall was approximated by a I-dimensional slab geometry with an 11/16" outer carbon steel wall, an inner stainless wall and 120 cm thick perlite zone. Runs were made for cases with expanded perlite, compacted perlite or with various void fractions. Runs were also made to simulate the effect of adding a moderator. Tallies were made for decay-time analysis from t=0 to 10 ms; total detected gamma

The purpose of this study is to develop a patient-specific dosimetry estimation system in nuclear medicine examination using a SimSET-based Monte Carlo code. We added a dose deposition routine to store the deposited energy of the photons during their flights in SimSET and developed a user-friendly interface for reading PET and CT images. Dose calculated on ORNL phantom was used to validate the accuracy of this system. The S values for {sup 99m}Tc, {sup 18}F and {sup 131}I obtained by the system were compared to those from the MCNP4C code and OLINDA. The ratios of S values computed by thismore » system to those obtained with OLINDA for various organs were ranged from 0.93 to 1.18, which are comparable to that obtained from MCNP4C code (0.94 to 1.20). The average ratios of S value were 0.99{+-}0.04, 1.03{+-}0.05, and 1.00{+-}0.07 for isotopes {sup 131}I, {sup 18}F, and {sup 99m}Tc, respectively. The simulation time of SimSET was two times faster than MCNP4C's for various isotopes. A 3D dose calculation was also performed on a patient data set with PET/CT examination using this system. Results from the patient data showed that the estimated S values using this system differed slightly from those of OLINDA for ORNL phantom. In conclusion, this system can generate patient-specific dose distribution and display the isodose curves on top of the anatomic structure through a friendly graphic user interface. It may also provide a useful tool to establish an appropriate dose-reduction strategy to patients in nuclear medicine environments. (authors)« less

The paper reports on a new utility for development of computational phantoms for Monte Carlo calculations and data analysis for in vivo measurements of radionuclides deposited in tissues. The individual properties of each worker can be acquired for a rather precise geometric representation of his (her) anatomy, which is particularly important for low energy gamma ray emitting sources such as thorium, uranium, plutonium and other actinides. The software discussed here enables automatic creation of an MCNP input data file based on scanning data. The utility includes segmentation of images obtained with either computed tomography or magnetic resonance imaging by distinguishing tissues according to their signal (brightness) and specification of the source and detector. In addition, a coupling of individual voxels within the tissue is used to reduce the memory demand and to increase the calculational speed. The utility was tested for low energy emitters in plastic and biological tissues as well as for computed tomography and magnetic resonance imaging scanning information.

The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

An X-ray control algorithm to modulate the X-ray intensity distribution over the FOV (field of view) has been developed by using numerical analysis and MCNP5, a particle transport simulation code on the basis of the Monte Carlo method. X-rays, which are widely used in medical diagnostic imaging, should be controlled in order to maximize the performance of the X-ray imaging system. However, transporting X-rays, like a liquid or a gas is conveyed through a physical form such as pipes, is not possible. In the present study, an X-ray control algorithm and technique to uniformize the Xray intensity projected on the image sensor were developed using a flattening filter and a collimator in order to alleviate the anisotropy of the distribution of X-rays due to intrinsic features of the X-ray generator. The proposed method, which is combined with MCNP5 modeling and numerical analysis, aimed to optimize a flattening filter and a collimator for a uniform distribution of X-rays. Their size and shape were estimated from the method. The simulation and the experimental results both showed that the method yielded an intensity distribution over an X-ray field of 6×4 cm2 at SID (source to image-receptor distance) of 5 cm with a uniformity of more than 90% when the flattening filter and the collimator were mounted on the system. The proposed algorithm and technique are not only confined to flattening filter development but can also be applied for other X-ray related research and development efforts.

University of Dallas (UD) physics majors are offered a variety of undergraduate research opportunities in nuclear physics through an established program at the University of Kentucky Accelerator Laboratory (UKAL). The 7-MV Model CN Van de Graaff accelerator and the neutron production and detection facilities located there are used by UD students to investigate how neutrons scatter from materials that are important in nuclear energy production and for our basic understanding of how neutrons interact with matter. Recent student projects include modeling of the laboratory using the neutron transport code MCNP to investigate the effectiveness of laboratory shielding, testing the long-termmore » gain stability of C 6D 6 liquid scintillation detectors, and deducing neutron elastic and inelastic scattering cross sections for 12C. Finally, results of these student projects are presented that indicate the pit below the scattering area reduces background by as much as 30%; the detectors show no significant gain instabilities; and new insights into existing 12C neutron inelastic scattering cross-section discrepancies near a neutron energy of 6.0 MeV are obtained.« less

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

University of Dallas (UD) physics majors are offered a variety of undergraduate research opportunities in nuclear physics through an established program at the University of Kentucky Accelerator Laboratory (UKAL). The 7-MV Model CN Van de Graaff accelerator and the neutron production and detection facilities located there are used by UD students to investigate how neutrons scatter from materials that are important in nuclear energy production and for our basic understanding of how neutrons interact with matter. Recent student projects include modeling of the laboratory using the neutron transport code MCNP to investigate the effectiveness of laboratory shielding, testing the long-termmore » gain stability of C 6D 6 liquid scintillation detectors, and deducing neutron elastic and inelastic scattering cross sections for 12C. Finally, results of these student projects are presented that indicate the pit below the scattering area reduces background by as much as 30%; the detectors show no significant gain instabilities; and new insights into existing 12C neutron inelastic scattering cross-section discrepancies near a neutron energy of 6.0 MeV are obtained.« less

This book contains part of the series New Liberal Arts, which is intended to make science and technology more accessible to students of the liberal arts. Volume in hand provides a comprehensive, multifaceted examination of nuclear energy, in nontechnical terms. Wolfson explains the basics of nuclear energy and radiation, nuclear power..., and nuclear weapons..., and he invites readers to make their own judgments on controversial nuclear issues. Illustrated with photos and diagrams. Each chapter contains suggestions for additional reading and a glossary. For policy, science, and general collections in all libraries. (ES) Topics contained include Atoms and nuclei. Effects andmore » uses of radiation. Energy and People. Reactor safety. Nuclear strategy. Defense in the nuclear age. Nuclear power, nuclear weapons, and nuclear futures.« less

Whenever nuclear material is found out of regulatory control, questions on the origin of the material, on its intended use, and on hazards associated with the material need to be answered. Analytical and interpretational methodologies have been developed in order to exploit measurable material properties for gaining information on the history of the nuclear material. This area of research is referred to as nuclear forensic science or, in short, nuclear forensics.This chapter reviews the origins, types, and state-of-the-art of nuclear forensics; discusses the potential roles of nuclear forensics in supporting nuclear security; and examines what nuclear forensics can realistically achieve.more » It also charts a path forward, pointing at potential applications of nuclear forensic methodologies in other areas.« less

Nuclear scans use radioactive substances to see structures and functions inside your body. They use a special ... images. Most scans take 20 to 45 minutes. Nuclear scans can help doctors diagnose many conditions, including ...

Trends in and factors related to the nuclear industry and nuclear fuel production are discussed. Topics addressed include nuclear reactors, survival of the U.S. uranium industry, production costs, budget cuts by the Department of Energy and U.S. Geological survey for resource studies, mining, and research/development activities. (JN)

Nuclear oil-well logging measures specific properties of subsurface geological formations as a function of depth in the well. The knowledge gained is used to evaluate the hydrocarbon potential of the surrounding oil field. The measurements are made by lowering an instrument package into an oil well and slowly extracting it at a constant speed. During the extraction phase, neutrons or gamma rays are emitted from the tool, interact with the formation, and scatter back to the detectors located within the tool. Even though only a small percentage of the emitted particles ever reach the detectors, mathematical modeling has been verymore » successful in the accurate prediction of these detector responses. The two dominant methods used to model these devices have been the two-dimensional discrete ordinates method and the three-dimensional Monte Carlo method has routinely been used to investigate the response characteristics of nuclear tools. A special Los Alamos National Laboratory version of their standard MCNP Monte carlo code retains the details of each particle history of later viewing within SABRINA, a companion three-dimensional geometry modeling and debugging code.« less

The spatial resolution of a detector is a very important parameter for x-ray imaging. A bulk scintillation detector because of spreading of light inside the scintillator does't have a good spatial resolution. The nanowire scintillators because of their wave guiding behavior can prevent the spreading of light and can improve the spatial resolution of traditional scintillation detectors. The zinc oxide (ZnO) scintillator nanowire, with its simple construction by electrochemical deposition in regular hexagonal structure of Aluminum oxide membrane has many advantages. The three dimensional absorption of X-ray energy in ZnO scintillator is simulated by a Monte Carlo transport code (MCNP). The transport, attenuation and scattering of the generated photons are simulated by a general-purpose scintillator light response simulation code (OPTICS). The results are compared with a previous publication which used a simulation code of the passage of particles through matter (Geant4). The results verify that this scintillator nanowire structure has a spatial resolution less than one micrometer.

The Health Protection Agency is changing from using detectors made from 7LiF:Mg,Ti in its photon/electron personal dosemeters, to 7LiF:Mg,Cu,P. Specifically, the Harshaw TLD-700H card is to be adopted. As a consequence of this change, the dosemeter holder is also being modified not only to accommodate the shape of the new card, but also to optimize the photon and electron response characteristics of the device. This redesign process was achieved using MCNP-4C2 and the kerma approximation, electron range/energy tables with additional electron transport calculations, and experimental validation, with different potential filters compared; the optimum filter studied was a polytetrafluoroethylene disc of diameter 18 mm and thickness 4.3 mm. Calculated relative response characteristics at different angles of incidence and energies between 16 and 6174 keV are presented for this new dosemeter configuration and compared with measured type-test results. A new estimate for the energy-dependent relative light conversion efficiency appropriate to the 7LiF:Mg,Cu,P was also derived for determining the correct dosemeter response.

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

The GATE Monte Carlo simulation platform based on the GEANT4 toolkit has come into widespread use for simulating positron emission tomography (PET) and single photon emission computed tomography (SPECT) imaging devices. Here, we explore its use for calculating electron dose distributions in water. Mono-energetic electron dose point kernels and pencil beam kernels in water are calculated for different energies between 15 keV and 20 MeV by means of GATE 6.0, which makes use of the GEANT4 version 9.2 Standard Electromagnetic Physics Package. The results are compared to the well-validated codes EGSnrc and MCNP4C. It is shown that recent improvements made to the GEANT4/GATE software result in significantly better agreement with the other codes. We furthermore illustrate several issues of general interest to GATE and GEANT4 users who wish to perform accurate simulations involving electrons. Provided that the electron step size is sufficiently restricted, GATE 6.0 and EGSnrc dose point kernels are shown to agree to within less than 3% of the maximum dose between 50 keV and 4 MeV, while pencil beam kernels are found to agree to within less than 4% of the maximum dose between 15 keV and 20 MeV.

This paper presents a validation test comparing angular distributions from an electron multiple-scattering experiment with those generated using the MCNP6 Monte Carlo code system. In this experiment, a 13- and 20-MeV electron pencil beam is deflected by thin foils with atomic numbers from 4 to 79. To determine the angular distribution, the fluence is measured down range of the scattering foil at various radii orthogonal to the beam line. The characteristic angle (the angle for which the max of the distribution is reduced by 1/e) is then determined from the angular distribution and compared with experiment. Multiple scattering foils tested herein include beryllium, carbon, aluminum, copper, and gold. For the default electron-photon transport settings, the calculated characteristic angle was statistically distinguishable from measurement and generally broader than the measured distributions. The average relative difference ranged from 5.8% to 12.2% over all of the foils, source energies, and physics settings tested. This validation illuminated a deficiency in the computation of the underlying angular distributions that is well understood. As a result, code enhancements were made to stabilize the angular distributions in the presence of very small substeps. However, the enhancement only marginally improved results indicating that additional algorithmic details should be studied.

Brachytherapy is a radiotherapy treatment where encapsulated radioactive sources are introduced within a patient. Depending on the technique used, such sources can produce high, medium or low local dose rates. The Monte Carlo method is a powerful tool to simulate sources and devices in order to help physicists in treatment planning. In multiple types of gynaecological cancer, intracavitary brachytherapy (HDR Ir-192 source) is used combined with other therapy treatment to give an additional local dose to the tumour. Different types of applicators are used in order to increase the dose imparted to the tumour and to limit the effect on healthy surrounding tissues. The aim of this work is to model both applicator and HDR source in order to evaluate the dose at a reference point as well as the effect of the materials constituting the applicators on the near field dose. The MCNP5 code based on the Monte Carlo method has been used for the simulation. Dose calculations have been performed with *F8 energy deposition tally, taking into account photons and electrons. Results from simulation have been compared with experimental in-phantom dose measurements. Differences between calculations and measurements are lower than 5%.The importance of the source position has been underlined.

Nuclear lamins are intermediate filament proteins that represent important structural components of metazoan nuclear envelopes (NEs). By combining proteomics and superresolution microscopy, we recently reported that both A- and B-type nuclear lamins form spatially distinct filament networks at the nuclear periphery of mouse fibroblasts. In particular, A-type lamins exhibit differential association with nuclear pore complexes (NPCs). Our studies reveal that the nuclear lamina network in mammalian somatic cells is less ordered and more complex than that of amphibian oocytes, the only other system in which the lamina has been visualized at high resolution. In addition, the NPC component Tpr likely links NPCs to the A-type lamin network, an association that appears to be regulated by C-terminal modification of various A-type lamin isoforms. Many questions remain, however, concerning the structure and assembly of lamin filaments, as well as with their mode of association with other nuclear components such as peripheral chromatin.

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

In the frame of the French trans-governmental R&D program against chemical, biological, radiological, nuclear and explosives (CBRN-E) threats, CEA is studying the detection of Special Nuclear Materials (SNM) by neutron interrogation with fast neutrons produced by an associated particle sealed tube neutron generator. The deuterium-tritium fusion reaction produces an alpha particle and a 14 MeV neutron almost back to back, allowing tagging neutron emission both in time and direction with an alpha particle position-sensitive sensor embedded in the generator. Fission prompt neutrons and gamma rays induced by tagged neutrons which are tagged by an alpha particle are detected in coincidence with plastic scintillators. This paper presents numerical simulations performed with the MCNP-PoliMi Monte Carlo computer code and with post processing software developed with the ROOT data analysis package. False coincidences due to neutron and photon scattering between adjacent detectors (cross talk) are filtered out to increase the selectivity between nuclear and benign materials. Accidental coincidences, which are not correlated to an alpha particle, are also taken into account in the numerical model, as well as counting statistics, and the time-energy resolution of the data acquisition system. Such realistic calculations show that relevant quantities of SNM (few kg) can be distinguished from cargo and shielding materials in 10 min acquisitions. First laboratory tests of the system under development in CEA laboratories are also presented.

This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

The nuclear lamins are type V intermediate filament proteins that are critically important for the structural properties of the nucleus. In addition, they are involved in the regulation of numerous nuclear processes, including DNA replication, transcription and chromatin organization. The developmentally regulated expression of lamins suggests that they are involved in cellular differentiation. Their assembly dynamic properties throughout the cell cycle, particularly in mitosis, are influenced by posttranslational modifications. Lamins may regulate nuclear functions by direct interactions with chromatin and determining the spatial organization of chromosomes within the nuclear space. They may also regulate chromatin functions by interacting with factors that epigenetically modify the chromatin or directly regulate replication or transcription. PMID:20826548

This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

A review of works performed at the Institute for Nuclear Research of the Russian Academy of Sciences and at the International Laser Center of the Moscow State University in the context of the new research area called "nuclear photonics" is presented. Nuclear photonics is based on creation of the new-generation gamma-ray sources which make it possible to solve a number of fundamental and applied problems, including research of low-energy photonuclear reactions, namely, investigation of collective excitations of nuclei near the threshold (pygmy resonances); nuclear safety assurance; production of low-energy positron beams; and phase-contrast X-ray imaging.

The purpose of this study was to compare and validate three methods to simulate radiographic image detectors with the Monte Carlo software MCNP/MCNPX in a time efficient way. The first detector model was the standard semideterministic radiography tally, which has been used in previous image simulation studies. Next to the radiography tally two alternative stochastic detector models were developed: A perfect energy integrating detector and a detector based on the energy absorbed in the detector material. Validation of three image detector models was performed by comparing calculated scatter-to-primary ratios (SPRs) with the published and experimentally acquired SPR values. For mammographic applications, SPRs computed with the radiography tally were up to 44% larger than the published results, while the SPRs computed with the perfect energy integrating detectors and the blur-free absorbed energy detector model were, on the average, 0.3% (ranging from -3% to 3%) and 0.4% (ranging from -5% to 5%) lower, respectively. For general radiography applications, the radiography tally overestimated the measured SPR by as much as 46%. The SPRs calculated with the perfect energy integrating detectors were, on the average, 4.7% (ranging from -5.3% to -4%) lower than the measured SPRs, whereas for the blur-free absorbed energy detector model, the calculated SPRs were, on the average, 1.3% (ranging from -0.1% to 2.4%) larger than the measured SPRs. For mammographic applications, both the perfect energy integrating detector model and the blur-free energy absorbing detector model can be used to simulate image detectors, whereas for conventional x-ray imaging using higher energies, the blur-free energy absorbing detector model is the most appropriate image detector model. The radiography tally overestimates the scattered part and should therefore not be used to simulate radiographic image detectors.

In order to detect Special Nuclear Materials (SNM) in unattended luggage or cargo containers in the field of homeland security, fissions are induced by 14 MeV neutrons produced by an associated particle DT neutron generator, and prompt fission particles correlated with tagged neutron are detected by plastic scintillators. SMN produce high multiplicity events due to induced fissions, whereas nonnuclear materials produce low multiplicity events due to cross-talk, (n,2n) or (n,n'γ) reactions. The data acquisition electronics is made of compact FPGA boards. The coincidence window is triggered by the alpha particle detection, allowing to tag the emission date and direction ofmore » the 14 MeV interrogating neutron. The first part of the paper presents experiment vs. calculation comparisons to validate MCNP-PoliMi simulations and the post-processing tools developed with the data analysis framework ROOT. Measurements have been performed using different targets (iron, lead, graphite), first with small plastic scintillators (10 x 10 x 10 cm{sup 3}) and then with large detectors (10 x 10 x 100 cm{sup 3}) to demonstrate that nuclear materials can be differentiated from nonnuclear dense materials (iron, lead) in iron and wood matrixes. A special attention is paid on SNM detection in abandoned luggage. In the second part of the paper, the performances of a cargo container inspection system are studied by numerical simulation, following previous work reported in. Detectors dimensions and shielding against the neutron generator background are optimized for container inspection. Events not correlated to an alpha particle (uncorrelated background), counting statistics, time and energy resolutions of the data acquisition system are all taken into account in a realistic numerical model. The impact of the container matrix (iron, ceramic, wood) has been investigated by studying the system capability to detect a few kilograms of SNM in different positions in the cargo container

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Nuclear Cardiology is a well-written, concise compendium of chapters that covers a wide range of topics in nuclear cardiology. Each chapter has been contributed by one or more recognized experts in the field, and the work is thoroughly referened. The physics and physiology of nuclear cardiology, myocardial imaging with /sup 201/Tl and /sup 99m/Tc pyrophosphate, left ventricular and right ventricular function, measurement of coronary blood flow with /sup 133/Xe, and microspheres are discussed, and there are chapters on metabolic imaging with positron emitters and on transmission computerized tomography of the heart.

The energy demand complexion of this country is always changing and promises to change in the future. The nuclear industry is responding to changing energy demands through standards writing activities. Since the oil embargo of 1973, there has been a change in the mix of fuels contributing to energy growth in this country; virtually all of the energy growth has come from coal and nuclear power. The predicted expansion of coal use by 1985, over 1977 level, is 37%, while the use of oil is expected to decline by 17%. Use of nuclear power is expected to increase 62% frommore » the 1977 level. The feasibility of using nuclear energy to meet the needs of the USA for electric power is discussed.« less

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Material about nuclear disarmament and the arms race should be included in secondary school curricula. Teachers can present this technical, controversial, and frightening material in a balanced and comprehensible way. Resources for instructional materials are listed. (PP)

Nanosecond Neutron Analysis / Associated Particles Technique (NNA/APT) has been used to create devices for detection of explosives, radioactive and heavily shielded nuclear materials in cargo containers. Explosives and other hazardous materials are detected by analyzing secondary high-energy gamma-rays form reactions of fast neutrons with the materials inside the container. Depending on the dimensions of the inspected containers, the detecting system consists of one or several detection modules, each of which contains a small neutron generator with built-in position sensitive detector of associated alpha-particles and several scintillator-based gamma-ray detectors. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. Array of several detectors of fast neutrons is used to detect neutrons from spontaneous and induced fission of nuclear materials. These neutrons can penetrate thick layers of lead shielding, which can be used to conceal gamma-radioactivity from nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 code was used to estimate the sensitivity of the device and its optimal configuration. Capability of the device to detect 1 kg of explosive imitator inside container filled with suitcases and other baggage items has been confirmed experimentally. First experiments with heavily shielded nuclear materials have been carried out.

The relationship between nuclear medicine and cardiology has continued to produce a surfeit of interesting, illuminating, and important reports involving the analysis of cardiac function, perfusion, and metabolism. To simplify the presentation, this review is broken down into three major subheadings: analysis of myocardial perfusion; imaging of the recent myocardial infarction; and the evaluation of myocardial function. There appears to be an increasingly important relationship between cardiology, particularly cardiac physiology, and nuclear imaging techniques. (KRM)

A safeguards neutron coincidence counter for nuclear material accountancy of the Advanced spent-fuel Conditioning Process Facility (ACPF), known as the ACP Safeguards Neutron Counter (ASNC), was upgraded to improve its remote-handling and maintenance capabilities. Based on the results of the previous design study, the neutron counter was completely rebuilt, and various detector parameters for neutron coincidence counting (i.e., high-voltage plateau, efficiency profile, dead time, die-away time, gate length, doubles gate fraction, and stability) were experimentally determined. The measurement data showed good agreement with the MCNP simulation results. To the best of the authors' knowledge, the ASNC is the only safeguards neutron coincidence counter in the world that is installed and operated in a hot-cell. The final goals to be achieved were (1) to evaluate the uncertainty level of the ASNC in nuclear material accountancy of the process materials of the oxide-reduction process for spent fuels and (2) to evaluate the applicability of the neutron coincidence counting technique within a strong radiation field (e.g., in a hot-cell environment).

Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work,more » we present the unfolding results using the EM algorithm.« less

Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted.

This table lists quantities of warheads (in stockpile, peak number per year, total number built, number of known test explosions), weapon development milestones (developers of the atomic bomb and hydrogen bomb, date of first operational ICBM, first nuclear-powered naval SSN in service, first MIRVed missile deployed), and testing milestones (first fission test, type of boosted fission weapon, multistage thermonuclear test, number of months from fission bomb to multistage thermonuclear bomb, etc.), and nuclear infrastructure (assembly plants, plutonium production reactors, uranium enrichment plants, etc.). Countries included in the tally are the United States, Soviet Union, Britain, France, and China.

Climate is universal. If a major modern nuclear war (i.e., with a large number of small-yield weapons) were to happen, it is not even necessary to have a specific part of the world directly involved for there to be cause to worry about the consequences for its inhabitants and their future. Indeed, smoke from fires ignited by the nuclear explosions would be transported by winds all over the world, causing dark and cold. According to the first study, by Turco et al. [1983], air surface temperature over continental areas of the northern mid-latitudes (assumed to be the nuclear war theatre) would fall to winter levels even in summer (hence the term “nuclear winter”) and induce drastic climatic conditions for several months at least. The devastating effects of a nuclear war would thus last much longer than was assumed initially. Discussing to what extent these estimations of long-term impacts on climate are reliable is the purpose of this article.

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞, macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

An analysis of the gamma-ray spectra produced using the quantum mechanical rotational energy formula is presented for nuclei with large angular momentum. This analysis is suitable for quantum mechanics, modern physics, or nuclear physics courses. (AIP) {copyright}{ital 1997 American Institute of Physics}

Pennsylvania State Dept. of Education, Harrisburg. Bureau of Curriculum Services.

This document is a report on a course in nuclear science for the high school curriculum. The course is designed to provide a basic but comprehensive understanding of the atom in the light of modern knowledge, and to show how people attempt to harness the tremendous energy liberated through fission and fusion reactions. The course crosses what are…

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Many scientists feel that research into nuclear safety has been diverted or distorted, and the results of the research concealed or inaccurately reported on a large number of occasions. Of particular concern have been the emergency cooling systems which have not, as yet, been adequately tested. (Author/MA)

ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Radioactive waste is mounting at U.S. nuclear power plants at a rate of more than 2,000 metric tons a year. Pursuant to statute and anticipating that a geologic repository would be available in 1998, the Department of Energy (DOE) entered into disposal contracts with nuclear utilities. Now, however, DOE does not expect the repository to be ready before 2010. For this reason, DOE does not want to develop a facility for monitored retrievable storage (MRS) by 1998. This book is concerned about how best to store the waste until a repository is available, congressional requesters asked GAO to review themore » alternatives of continued storage at utilities' reactor sites or transferring waste to an MRS facility, GAO assessed the likelihood of an MRSA facility operating by 1998, legal implications if DOE is not able to take delivery of wastes in 1998, propriety of using the Nuclear Waste Fund-from which DOE's waste program costs are paid-to pay utilities for on-site storage capacity added after 1998, ability of utilities to store their waste on-site until a repository is operating, and relative costs and safety of the two storage alternatives.« less

Explains problems enforcing the Nuclear Non-Proliferation Treaty (NPT) of 1968. Provides factual charts and details concerning the production of nuclear energy and arms, the processing and disposal of waste products, and outlines the nuclear fuel cycle. Discusses safeguards, the risk of nuclear terrorism, and ways to deal with these problems. (NL)

Current world stockpiles of nuclear weapons and the status of treaties for nuclear disarmament and the ultimate elimination of nuclear weapons are summarised. The need for including stockpiles of civil plutonium in a programme for ending production and disposing of fissile materials is emphasized, and the ultimate difficulty of disposing of the last few nuclear weapons discussed.

Once they have nuclear power, most countries will divert nuclear materials from commercial to military programs. In excerpts from the book "Facing the Danger" (by Totten, S. and M. W., Crossing Press, 1984), five anti-nuclear activists explain how and why they have been addressing the nuclear connection. (RM)

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

Calculations of criticality of the AP1000 core due to the use of new edition of nuclear data library namely ENDF/B-VII and ENDF/B-VII.1 have been done. This work is aimed to know the accuracy of ENDF/B-VII.1 compared to ENDF/B-VII and ENDF/B-VI.8. in determining the criticality parameter of AP1000. Analysis ws imposed to core at cold zero power (CZP) conditions. The calculations have been carried out by means of MCNP computer code for 3 dimension geometry. The results show that criticality parameter namely effective multiplication factor of the AP1000 core are higher than that ones resulted from ENDF/B-VI.8 with relative differences of 0.39% for application of ENDF/B-VII and of 0.34% for application of ENDF/B-VII.1.

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applications of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applicationsmore » of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Lastly, each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.« less

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applications of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applicationsmore » of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Lastly, each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.« less

calorimeter were carried out. A preliminary analysis shows that the numerical results overestimate the measurements by about 20 %. A new approach has been developed in order to estimate the nuclear heating by two methods (energy deposition or KERMA) by considering the whole complete geometry of the sensor. This new approach will contribute to the interpretation of the irradiation campaign and will be useful to improve the out-of-pile calibration procedure of the sensor and its thermal response during irradiations. The aim of this paper is to present simulations made by using MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) for the nuclear heating inside the different parts of the calorimeter (head, rod and base). Calculations into two steps will be realized. We will use as an input source in the model new spectra (neutrons, prompt-photons and delayed-photons) calculated with the Monte Carlo code TRIPOLI-4{sup R} inside different experimental channels (water) located into the OSIRIS periphery and used during the CARMEN-1P irradiation campaign. We will consider Neutrons- Photons-Electrons and Photons-Electrons modes. We will begin by a brief description of the differential-calorimeter device geometry. Then the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements will be introduced. The energy deposition due to the prompt-gamma, delayed-gamma and neutrons, the neutron-activation of the device will be considered. The different components of the nuclear heating inside the different parts of the calorimeter will be detailed. Moreover, a comparison between KERMA and nuclear energy deposition estimations will be given. Finally, a comparison between this total nuclear heating Calculation and Experiment in graphite sample will be determined. (authors)« less

A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 70 cm3) aimed at testing and validating the recent nuclear data libraries for fusion applications was performed in the frame of the European Fusion Program at the 14 MeV ENEA Frascati Neutron Generator (FNG). Reaction rates, neutron flux spectra and doses were measured using different experimental techniques (e.g. activation foils techniques, NE213 scintillator and thermoluminescent detectors). This paper first summarizes the analyses of the experiment carried-out using the MCNP5 Monte Carlo code and the European JEFF-3.2 library. Large discrepancies between calculation (C) and experiment (E) were found for the reaction rates both in the high and low neutron energy range. The analysis was complemented by sensitivity/uncertainty analyses (S/U) using the deterministic and Monte Carlo SUSD3D and MCSEN codes, respectively. The S/U analyses enabled to identify the cross sections and energy ranges which are mostly affecting the calculated responses. The largest discrepancy among the C/E values was observed for the thermal (capture) reactions indicating severe deficiencies in the 63,65Cu capture and elastic cross sections at lower rather than at high energy. Deterministic and MC codes produced similar results. The 14 MeV copper experiment and its analysis thus calls for a revision of the JEFF-3.2 copper cross section and covariance data evaluation. A new analysis of the experiment was performed with the MCNP5 code using the revised JEFF-3.3-T2 library released by NEA and a new, not yet distributed, revised JEFF-3.2 Cu evaluation produced by KIT. A noticeable improvement of the C/E results was obtained with both new libraries.

A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

Nuclear energy can play a role in carbon free production of electrical energy, thus making it interesting for tomorrow's energy mix. However, several issues have to be addressed. In fission technology, the design of so-called fourth generation reactors show great promise, in particular in addressing materials efficiency and safety issues. If successfully developed, such reactors may have an important and sustainable part in future energy production. Working fusion reactors may be even more materials efficient and environmental friendly, but also need more development and research. The roadmap for development of fourth generation fission and fusion reactors, therefore, asks for attention and research in these fields must be strengthened.

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

In case of a terrorist nuclear attack in a metropolitan area, EMP measurement could provide: (1) a prompt confirmation of the nature of the explosion (chemical or nuclear) for emergency response; and (2) and characterization parameters of the device (reaction history, yield) for technical forensics. However, urban environment could affect the fidelity of the prompt EMP measurement (as well as all other types of prompt measurement): (1) Nuclear EMP wavefront would no longer be coherent, due to incoherent production, attenuation, and propagation of gamma and electrons; and (2) EMP propagation from source region outward would undergo complicated transmission, reflection, andmore » diffraction processes. EMP simulation for electrically-large urban environment: (1) Coupled MCNP/FDTD (Finite-difference time domain Maxwell solver) approach; and (2) FDTD tends to be limited to problems that are not 'too' large compared to the wavelengths of interest because of numerical dispersion and anisotropy. We use a higher-order low-dispersion, isotropic FDTD algorithm for EMP propagation.« less

The thermal neutron-elastic-scattering cross-section data for {sup 16}O used in various modern evaluated-nuclear-data libraries were reviewed and found to be generally too high compared with the best available experimental measurements. Some of the proposed revisions to the ENDF/B-VII.0 {sup 16}O data library and recent results from the TENDL system increase this discrepancy further. The reactivity impact of revising the {sup 16}O data downward to be consistent with the best measurements was tested using the JENDL-3.3 {sup 16}O cross-section values and was found to be very small in MCNP5 simulations of the UO{sub 2} and reactor-recycle MOX-fuel cases of the ANSmore » Doppler-defect numerical benchmark. However, large reactivity differences of up to about 14 mk (1400 pcm) were observed using {sup 16}O data files from several evaluated-nuclear-data libraries in MCNP5 simulations of the Los Alamos National Laboratory HEU heavy-water solution thermal critical experiments, which were performed in the 1950's. The latter result suggests that new measurements using HEU in a heavy-water-moderated critical facility, such as the ZED-2 zero-power reactor at the Chalk River Laboratories, might help to resolve the discrepancy between the {sup 16}O thermal elastic-scattering cross-section values and thereby reduce or better define its uncertainty, although additional assessment work would be needed to confirm this. (authors)« less

with foreign companies, will be the first country’s nuclear arsenal. large nuclear power plant in China when completed. It researched and designed...JPRS-TND-89-022 29 NOVEMBER 1989, FOREIGN BROADCAST INFORMATION SERVICE -IPRS Report- Nuclear Developments REPRODUCED BY U.S. DEPARTMENT OF COMMERCE...1 Commission Views West Coast for Nuclear Sites [SAPA] ................................................................. I CHINA Nuclear

The Department of Energy's (DOE) Lawrence Livermore National Laboratory, located in Livermore, California, generates and controls large numbers of classified documents associated with the research and testing of nuclear weapons. Concern has been raised about the potential for espionage at the laboratory and the national security implications of classified documents being stolen. This paper determines the extent of missing classified documents at the laboratory and assesses the adequacy of accountability over classified documents in the laboratory's custody. Audit coverage was limited to the approximately 600,000 secret documents in the laboratory's custody. The adequacy of DOE's oversight of the laboratory's secretmore » document control program was also assessed.« less

A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Extreme states of nuclearmatter (such that feature high spins, large deformations, high density and temperature, or a large excess of neutrons and protons) play an important role in studying fundamental properties of nuclei and are helpful in solving the problem of constructing the equation of state for nuclear matter. The synthesis of neutron-rich nuclei near the nucleon drip lines and investigation of their properties permit drawing conclusions about the positions of these boundaries and deducing information about unusual states of such nuclei and about their decays. At the present time, experimental investigations along these lines can only be performed viamore » the cooperation of leading research centers that possess powerful heavy-ion accelerators, such as the Large Hadron Collider (LHC) at CERN and the heavy-ion cyclotrons at the Joint Institute for Nuclear Research (JINR, Dubna), where respective experiments are being conducted by physicists from about 20 JINR member countries. The present article gives a survey of the most recent results in the realms of super neutron-rich nuclei. Implications of the change in the structure of such nuclei near the nucleon drip lines are discussed. Information about the results obtained by measuring the masses (binding energies) of exotic nuclei, the nucleon-distribution radii (neutron halo) and momentum distributions in them, and their deformations and quantum properties is presented. It is shown that the properties of nuclei lying near the stability boundaries differ strongly from the properties of other nuclei. The problem of the stability of nuclei that is associated with the magic numbers of 20 and 28 is discussed along with the effect of new magic numbers.« less

Describes a simple model, based on nuclear shell considerations, which leads to the proper behavior of known nuclear quadrupole moments, although predictions of the magnitudes of some quadrupole moments are seriously in error.

This article reviews the nuclear weapons modernization programs underway in the world's nine nuclear weapons states. It concludes that despite significant reductions in overall weapons inventories since the end of the Cold War, the pace of reductions is slowing - four of the nuclear weapons states are even increasing their arsenals, and all the nuclear weapons states are busy modernizing their remaining arsenals in what appears to be a dynamic and counterproductive nuclear competition. The author questions whether perpetual modernization combined with no specific plan for the elimination of nuclear weapons is consistent with the nuclear Non-Proliferation Treaty and concludes that new limits on nuclear modernizations are needed.

In order to increase the efficiency of Monte Carlo simulation for nuclear well logging problems, a new method has been developed for variance reduction. With this method, an importance map is generated in the regular Monte Carlo calculation as a by-product, and the importance map is later used to conduct the splitting and Russian roulette for particle population control. By adopting a spatial mesh system, which is independent of physical geometrical configuration, the method allows superior user-friendliness. This new method is incorporated into the general purpose Monte Carlo code MCNP4A through a patch file. Two nuclear well logging problems, a neutron porosity tool and a gamma-ray lithology density tool are used to test the performance of this new method. The calculations are sped up over analog simulation by 120 and 2600 times, for the neutron porosity tool and for the gamma-ray lithology density log, respectively. The new method enjoys better performance by a factor of 4~6 times than that of MCNP's cell-based weight window, as per the converged figure-of-merits. An indirect comparison indicates that the new method also outperforms the AVATAR process for gamma-ray density tool problems. Even though it takes quite some time to generate a reasonable importance map from an analog run, a good initial map can create significant CPU time savings. This makes the method especially suitable for nuclear well logging problems, since one or several reference importance maps are usually available for a given tool. Study shows that the spatial mesh sizes should be chosen according to the mean-free-path. The overhead of the importance map generator is 6% and 14% for neutron and gamma-ray cases. The learning ability towards a correct importance map is also demonstrated. Although false-learning may happen, physical judgement can help diagnose with contributon maps. Calibration and analysis are performed for the neutron tool and the gamma-ray tool. Due to the fact that a very

Purpose: Validate implementation of a published RBE model for DSB induction (RBEDSB) in several general purpose Monte Carlo (MC) code systems and the RayStation™ treatment planning system (TPS). For protons and other light ions, DSB induction is a critical initiating molecular event that correlates well with the RBE for cell survival. Methods: An efficient algorithm to incorporate information on proton and light ion RBEDSB from the independently tested Monte Carlo Damage Simulation (MCDS) has now been integrated into MCNP (Stewart et al. PMB 60, 8249–8274, 2015), FLUKA, TOPAS and a research build of the RayStation™ TPS. To cross-validate the RBEDSBmore » model implementation LET distributions, depth-dose and lateral (dose and RBEDSB) profiles for monodirectional monoenergetic (100 to 200 MeV) protons incident on a water phantom are compared. The effects of recoil and secondary ion production ({sub 2}H{sub +}, {sub 3}H{sub +}, {sub 3}He{sub 2+}, {sub 4}He{sub 2+}), spot size (3 and 10 mm), and transport physics on beam profiles and RBEDSB are examined. Results: Depth-dose and RBEDSB profiles among all of the MC models are in excellent agreement using a 1 mm distance criterion (width of a voxel). For a 100 MeV proton beam (10 mm spot), RBEDSB = 1.2 ± 0.03 (− 2–3%) at the tip of the Bragg peak and increases to 1.59 ± 0.3 two mm distal to the Bragg peak. RBEDSB tends to decrease as the kinetic energy of the incident proton increases. Conclusion: The model for proton RBEDSB has been accurately implemented into FLUKA, MCNP, TOPAS and the RayStation™TPS. The transport of secondary light ions (Z > 1) has a significant impact on RBEDSB, especially distal to the Bragg peak, although light ions have a small effect on (dosexRBEDSB) profiles. The ability to incorporate spatial variations in proton RBE within a TPS creates new opportunities to individualize treatment plans and increase the therapeutic ratio. Dr. Erik Traneus is employed full-time as a Research

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

The state-of-the-art of the still-conceptual nuclear air cushion vehicle, particularly the nuclear powerplant is identified. Using mission studies and cost estimates, some of the advantages of nuclear power for large air cushion vehicles are described. The technology studies on mobile nuclear powerplants and conceptual ACV systems/missions studies are summarized.

The detection and characterization of special nuclear materials is a high priority area for applications in nuclear safeguards and nonproliferation. We are developing new instruments based on organic scintillators to detect and characterize the emissions from special nuclear materials. This paper describes some of the gaps and challenges in nuclear safeguards and proposed approaches.

This essay explores the ways terrorist groups may gain possession of nuclear materials; the way in which they may use nuclear weapons and other nuclear technologies to their benefit; and various courses of action designed to minimize the possibilities of terrorists utilizing nuclear technology to their benefit and society's detriment. (BT)

A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

In order to evaluate the neutron doses around nuclear fissile objects, a comparative study has been made on several neutron dosemeters: bubble dosemeters, etched-track detectors (CR-39) and 3He-filled proportional counters used as dose-rate meters. The measurements were made on the ambient and the personal dose equivalents H*(10) and Hp(10). Results showed that several bubble dosemeters should have been used due to a low reproducibility in the measurements. A strong correlation with the neutron energy was also found, with about a 30% underestimation of Hp(10) for neutrons from the PuBe source, and about a 9% overestimation for neutrons from the 252Cf source. Measurements of the nuclear fissile objects were made using the CR-39 and the dose-rate meters. The CR-39 led to an underestimation of 30% with respect to the neutron dose-rate meter measurements. In addition, the MCNP calculation code was used in the different configurations.

The size of hot particles contained in nuclear fallout ranges from 10 nm to 20 microm for the worldwide weapons fallout. Hot particles from nuclear power reactors can be significantly bigger (100 microm to several millimetres). Electron backscattering from such particles is a prominent secondary effect in beta dosimetry for radiological protection purposes, such as skin dosimetry. In this study, the effect of electron backscattering due to hot particles contamination on skin dose is investigated. These include parameters such as detector area, source radius, source energy, scattering material and source density. The Monte-Carlo Neutron Particle code (MCNP4C) was used to calculate the depth dose distribution for 10 different beta sources and various materials. The backscattering dose factors (BSDF) were then calculated. A significant dependence is shown for the BSDF magnitude upon detector area, source radius and scatterers. It is clearly shown that the BSDF increases with increasing detector area. For high Z scatterers, the BSDF can reach as high as 40 and 100% for sources with radii 0.1 and 0.0001 cm, respectively. The variation of BSDF with source radius, source energy and source density is discussed.

There is currently a great deal of technical and political effort focused on reducing the risk of potential attacks on the United States involving radiological dispersal devices or nuclear weapons. This paper proposes a benchmark problem for gamma-ray and X-ray cargo monitoring with results calculated using MCNP5, v1.51. The primary goal is to provide a benchmark problem that will allow researchers in this area to evaluate Monte Carlo models for both speed and accuracy in both forward and inverse calculational codes and approaches for nuclear security applications. A previous benchmark problem was developed by one of the authors (RPG) for two similar oil well logging problems (Gardner and Verghese, 1991, [1]). One of those benchmarks has recently been used by at least two researchers in the nuclear threat area to evaluate the speed and accuracy of Monte Carlo codes combined with variance reduction techniques. This apparent need has prompted us to design this benchmark problem specifically for the nuclear threat researcher. This benchmark consists of conceptual design and preliminary calculational results using gamma-ray interactions on a system containing three thicknesses of three different shielding materials. A point source is placed inside the three materials lead, aluminum, and plywood. The first two materials are in right circular cylindrical form while the third is a cube. The entire system rests on a sufficiently thick lead base so as to reduce undesired scattering events. The configuration was arranged in such a manner that as gamma-ray moves from the source outward it first passes through the lead circular cylinder, then the aluminum circular cylinder, and finally the wooden cube before reaching the detector. A 2 in.×4 in.×16 in. box style NaI (Tl) detector was placed 1 m from the point source located in the center with the 4 in.×16 in. side facing the system. The two sources used in the benchmark are 137Cs and 235U.

All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

The MIT Linear Electrostatic Ion Accelerator (LEIA) generates DD and D{sup 3}He fusion products for the development of nuclear diagnostics for Omega, Z, and the National Ignition Facility (NIF). Significant improvements to the system in recent years are presented. Fusion reaction rates, as high as 10{sup 7} s{sup -1} and 10{sup 6} s{sup -1} for DD and D{sup 3}He, respectively, are now well regulated with a new ion source and electronic gas control system. Charged fusion products are more accurately characterized, which allows for better calibration of existing nuclear diagnostics. In addition, in situ measurements of the on-target beam profile,more » made with a CCD camera, are used to determine the metrology of the fusion-product source for particle-counting applications. Finally, neutron diagnostics development has been facilitated by detailed Monte Carlo N-Particle Transport (MCNP) modeling of neutrons in the accelerator target chamber, which is used to correct for scattering within the system. These recent improvements have resulted in a versatile platform, which continues to support the existing nuclear diagnostics while simultaneously facilitating the development of new diagnostics in aid of the National Ignition Campaign at the National Ignition Facility.« less

Complex and exotic nuclear geometries, collectively referred to as "nuclear pasta," are expected to appear naturally in dense nuclear matter found in the crusts of neutron stars and supernovae environments. The pasta geometries depend on the average baryon density, proton fraction, and temperature and are critically important in the determination of many transport properties of matter in supernovae and the crusts of neutron stars. Using a set of self-consistent microscopic nuclear energy density functionals, we present the first results of large scale quantum simulations of pasta phases at baryon densities 0.03 ≤ρ ≤0.10 fm-3 , proton fractions 0.05 ≤Yp≤0.40 , and zero temperature. The full quantum simulations, in particular, allow us to thoroughly investigate the role and impact of the nuclear symmetry energy on pasta configurations. We use the Sky3D code that solves the Skyrme Hartree-Fock equations on a three-dimensional Cartesian grid. For the nuclear interaction we use the state-of-the-art UNEDF1 parametrization, which was introduced to study largely deformed nuclei, hence is suitable for studies of the nuclear pasta. Density dependence of the nuclear symmetry energy is simulated by tuning two purely isovector observables that are insensitive to the current available experimental data. We find that a minimum total number of nucleons A =2000 is necessary to prevent the results from containing spurious shell effects and to minimize finite size effects. We find that a variety of nuclear pasta geometries are present in the neutron star crust, and the result strongly depends on the nuclear symmetry energy. The impact of the nuclear symmetry energy is less pronounced as the proton fractions increase. Quantum nuclear pasta calculations at T =0 MeV are shown to get easily trapped in metastable states, and possible remedies to avoid metastable solutions are discussed.

Nuclear Astrophysics as a broad and diverse field of study can be viewed as a magnifier of the impact of microscopic processes on the evolution of macroscopic events. One of the primary goals in Nuclear Astrophysics is the understanding of the nucleosynthesis processes that take place in the cosmos and the simulation of the correlated stellar and explosive burning scenarios. These simulations are strongly dependent on the input from Nuclear Physics which sets the time scale for all stellar dynamic processes--from giga-years of stellar evolution to milliseconds of stellar explosions--and provides the basis for most of the signatures that wemore » have for the interpretation of these events--from stellar luminosities, elemental and isotopic abundances to neutrino flux from distant supernovae. The Nuclear Physics input comes through nuclear structure, low energy reaction rates, nuclear masses, and decay rates. There is a common perception that low energy reaction rates are the most important component of the required nuclear physics input; however, in this article we take a broader approach and present an overview of the close correlation between various nuclear structure aspects and their impact on nuclear astrophysics. We discuss the interplay between the weak and the strong forces on stellar time scales due to the limitations they provide for the evolution of slow and rapid burning processes. The effects of shell structure in nuclei on stellar burning processes as well as the impact of clustering in nuclei is outlined. Furthermore we illustrate the effects of the various nuclear structure aspects on the major nucleosynthesis processes that have been identified in the last few decades. We summarize and provide a coherent overview of the impact of all aspects of nuclear structure on nuclear astrophysics.« less

Whenever nuclear material is found out of regulatory control, questions on the origin of the material, on its intended use, and on hazards associated with the material need to be answered. Here, analytical and interpretational methodologies have been developed in order to exploit measurable material properties for gaining information on the history of the nuclear material. This area of research is referred to as nuclear forensic science or, in short, nuclear forensics.This chapter reviews the origins, types, and state-of-the-art of nuclear forensics; discusses the potential roles of nuclear forensics in supporting nuclear security; and examines what nuclear forensics can realisticallymore » achieve. Lastly, it also charts a path forward, pointing at potential applications of nuclear forensic methodologies in other areas.« less

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Recently there has been a lot of interest in small modular reactors. A specific type of these small modular reactors (SMR,) are marine based power plants called floating nuclear power plants (FNPP). These FNPPs are typically built by countries with extensive knowledge of nuclear energy, such as Russia, France, China and the US. These FNPPs are built in one country and then sent to countries in need of power and/or seawater desalination. Fifteen countries have expressed interest in acquiring such power stations. Some designs for such power stations are briefly summarized. Several different avenues for cooperation in FNPP technology aremore » proposed, including IAEA nuclear security (i.e. safeguards), multilateral or bilateral agreements, and working with Russian design that incorporates nuclear safeguards for IAEA inspections in non-nuclear weapons states« less

A nuclear radiation actuated valve for a nuclear reactor. The valve has a valve first part (such as a valve rod with piston) and a valve second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics. The valve's first part is positioned to receive nuclear radiation from the nuclear reactor's fuel region. The valve's second part is positioned so that its nuclear radiation induced swelling is different from that of the valve's first part. The valve's second part also is positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system.

This article reviews the nuclear weapons modernization programs underway in the world's nine nuclear weapons states. It concludes that despite significant reductions in overall weapons inventories since the end of the Cold War, the pace of reductions is slowing - four of the nuclear weapons states are even increasing their arsenals, and all the nuclear weapons states are busy modernizing their remaining arsenals in what appears to be a dynamic and counterproductive nuclear competition. The author questions whether perpetual modernization combined with no specific plan for the elimination of nuclear weapons is consistent with the nuclear Non-Proliferation Treaty and concludesmore » that new limits on nuclear modernizations are needed.« less

... page: //medlineplus.gov/ency/article/003835.htm RBC nuclear scan To use the sharing features on this page, please enable JavaScript. An RBC nuclear scan uses small amounts of radioactive material to ...

In 1988 the science historian Spencer Weart published a groundbreaking book called Nuclear Fear: A History of Images, which examined visions of radiation damage and nuclear disaster in newspapers, television, film, literature, advertisements and popular culture.

This report contains brief papers that discusses the following topics: Fundamental Symmetries in the Nucleus; Internucleon Interactions; Dynamics of Very Light Nuclei; Facets of the Nuclear Many-Body Problem; and Nuclear Instruments and Methods.

This paper reports that using Puget Sound Naval Shipyard as a model, GAO examined the Navy's accounting practices at nuclear shipyards. In fiscal year 1991, Puget Sound worked on 24 nuclear-powered and three conventionally powered ships. About 31 percent of the workdays and 35 percent of total costs were for nuclear work. The average cost per workday for nuclear labor was 25 percent higher than for non-nuclear work, and the average cost per day for overhead for nuclear work was about 60 percent higher. These higher costs are due to the complexity of nuclear work, which requires a higher levelmore » of services, and the higher cost of specially trained workers and specialized shipyard departments that support nuclear work.« less

Contends that courses on nuclear war must help students examine the political, social, religious, philosophical, economic, and moral assumptions which characterized the dilemma of nuclear armament/disarmament. Describes the upper level undergraduate course taught by the authors. (JDH)

arranged according to degree of severity, and action plans will be developed for necessary measures appropriate to each level. Nuclear Disaster Chief...For the purpose of shielding the country as much as possible from the damages of nuclear accidents, first, a " Nuclear Disaster Headquarters" will be...formed within TAEC to oversee operations called for by the " Nuclear Disaster Guidelines," and coordination of the activities designed to meet an

enrichment facility); 3. The acquisition of the technology and know-how to design, assemble, and manufacture the bomb ; 4. A full-scale nuclear test...14 Scott D. Sagan, “Why Do States Build Nuclear Weapons?: Three Models in Search of a Bomb ,” International...15 Sagan, “Why Do States Build Nuclear Weapons?: Three Models in Search of a Bomb ,” 57–59. 16 Lewis A. Dunn and Herman Kahn, Trends in Nuclear

Nuclear power is an important and, the authors believe, essential component of a secure nuclear future. Although nuclear fuel cycles create materials that have some potential for use in nuclear weapons, with appropriate fuel cycles, nuclear power could reduce rather than increase real proliferation risk worldwide. Future fuel cycles could be designed to avoid plutonium production, generate minimal amounts of plutonium in proliferation-resistant amounts or configurations, and/or transparently and efficiently consume plutonium already created. Furthermore, a strong and viable US nuclear infrastructure, of which nuclear power is a large element, is essential if the US is to maintain a leadershipmore » or even participatory role in defining the global nuclear infrastructure and controlling the proliferation of nuclear weapons. By focusing on new fuel cycles and new reactor technologies, it is possible to advantageously burn and reduce nuclear materials that could be used for nuclear weapons rather than increase and/or dispose of these materials. Thus, the authors suggest that planners for a secure nuclear future use technology to design an ideal future. In this future, nuclear power creates large amounts of virtually atmospherically clean energy while significantly lowering the threat of proliferation through the thoughtful use, physical security, and agreed-upon transparency of nuclear materials. The authors must develop options for policy makers that bring them as close as practical to this ideal. Just as Atoms for Peace became the ideal for the first nuclear century, they see a potential nuclear future that contributes significantly to power for peace and prosperity.« less

Nuclear Waste Storage Act of 2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage ...enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons materials...that had used the leased fuel , along with supplies of fresh nuclear fuel , according to the GNEP concept; see [http://www.gnep.energy.gov].

Fundamental principles governing nuclear explosions and their effects are discussed, including three components of a nuclear explosion (thermal radiation, shock wave, nuclear radiation). Describes how effects of these components depend on the weapon's yield, its height of burst, and distance of detonation point. Includes effects of three…

Basic concepts of nuclear structures, radiation, nuclear reactions, and health physics are presented in this text, prepared for naval officers. Applications to the area of nuclear power are described in connection with pressurized water reactors, experimental boiling water reactors, homogeneous reactor experiments, and experimental breeder…

Presents results of a survey of the teaching about nuclear history at U.S. colleges and universities. Reports the existence of a well-established and extensive literature, a focus on nuclear weapons or warfare, and a concentration on nuclear citizenship, therapy, or eschatology for courses outside of history departments. Discusses individual…

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

An overview of space nuclear energy technologies is presented. The development and characteristics of radioisotope thermoelectric generators (RTG's) and space nuclear power reactors are discussed. In addition, the policy and issues related to public safety and the use of nuclear power sources in space are addressed.

This software is code related to reading/writing/manipulating nuclear data in the Generalized Nuclear Data (GND) format, a new format for sharing nuclear data among institutions. In addition to the software and its documentation, notes and documentation from the WPEC Subgroup 43 will be included. WPEC Subgroup 43 is an international committee charged with creating the API for the GND format.

National Academy of Sciences - National Research Council, Washington, DC.

This report covers the general issues involved in nuclear safety research and points out the areas needing detailed consideration. Topics included are: (1) "Principles of Nuclear Safety Research" (examining who should fund, who should conduct, and who should set the agenda for nuclear safety research); (2) "Elements of a Future…

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. In NASA’s recent Mars Design Reference Architecture (DRA) 5.0 study (NASA-SP-2009-566, July 2009), nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option because of its high thrust and high specific impulse (-900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. An extensive nuclear thermal rocket technology development effortmore » was conducted from 1955-1973 under the Rover/NERVA Program. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art design incorporating lessons learned from the very successful technology development program. Past activities at the NASA Glenn Research Center have included development of highly detailed MCNP Monte Carlo transport models of the SNRE and other small engine designs. Preliminary core configurations typically employ fuel elements with fixed fuel composition and fissile material enrichment. Uniform fuel loadings result in undesirable radial power and temperature profiles in the engines. Engine performance can be improved by some combination of propellant flow control at the fuel element level and by varying the fuel composition. Enrichment zoning at the fuel element level with lower enrichments in the higher power elements at the core center and on the core periphery is particularly effective. Power flattening by enrichment zoning typically results in more uniform propellant exit temperatures and improved engine performance. For the SNRE, element enrichment zoning provided very flat radial power profiles with 551 of

Spent nuclear fuel data are collected by the U.S. Energy Information Administration (EIA) for the Department of Energy's Office of Standard Contract Management (Office of the General Counsel) on the Form GC-859, "Nuclear Fuel Data Survey." The data include detailed characteristics of spent nuclear fuel discharged from commercial U.S. nuclear power plants and currently stored at commercial sites in the United States. Utilities were not required to report spent nuclear fuel assemblies shipped to away-from-reactor, off-site facilities.

India emerged as a free and democratic country in 1947, and entered into the nuclear age in 1948 by establishing the Atomic Energy Commission (AEC), with Homi Bhabha as the chairman. Later on the Department of Atomic Energy (DAE) was created under the Office of the Prime Minister Jawahar Lal Nehru. Initially the AEC and DAE received international cooperation, and by 1963 India had two research reactors and four nuclear power reactors. In spite of the humiliating defeat in the border war by China in 1962 and China's nuclear testing in 1964, India continued to adhere to the peaceful uses of nuclear energy. On May 18, 1974 India performed a 15 kt Peaceful Nuclear Explosion (PNE). The western powers considered it nuclear weapons proliferation and cut off all financial and technical help, even for the production of nuclear power. However, India used existing infrastructure to build nuclear power reactors and exploded both fission and fusion devices on May 11 and 13, 1998. The international community viewed the later activity as a serious road block for the Non-Proliferation Treaty and the Comprehensive Test Ban Treaty; both deemed essential to stop the spread of nuclear weapons. India considers these treaties favoring nuclear states and is prepared to sign if genuine nuclear disarmament is included as an integral part of these treaties.

Concerns around global warming have led to a nuclear renaissance in many countries, meanwhile the nuclear industry is warning already of a need to train more nuclear engineers and scientists, who are needed in a range of areas from healthcare and radiation detection to space exploration and advanced materials as well as for the nuclear power industry. Here Karl Whittle provides a solid overview of the intersection of nuclear engineering and materials science at a level approachable by advanced students from materials, engineering and physics. The text explains the unique aspects needed in the design and implementation of materials for use in demanding nuclear settings. In addition to material properties and their interaction with radiation the book covers a range of topics including reactor design, fuels, fusion, future technologies and lessons learned from past incidents. Accompanied by problems, videos and teaching aids the book is suitable for a course text in nuclear materials and a reference for those already working in the field.

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

The management of radioactive waste ensures safety to human health and the environment nowadays and for the future, without overwhelming the upcoming generations. The primary characterization of radioactive waste is one of the main steps in the management of radioactive waste. This step permits to choose the best treatment for the radioactive waste before forwarding it to its final disposal. The aim of the present work is the primary characterization of cartridge filters from the IEA-R1 nuclear reactor utilizing gamma-ray spectrometry, and the method of Monte Carlo for calibration. The IEA-R1 is located in the Nuclear and Energy Research Institutemore » (IPEN - CNEN) in the city of Sao Paulo, Brazil. Cartridge filters are used for purification of the cooling water that is pumped through the core of the pool type nuclear research reactors. Once worn out, these filters are replaced and then become radioactive waste. Determination of the radioactive inventory is of paramount importance in the management of such radioactive waste, and one of the main methods for doing so is the gamma-ray spectrometry, which can identify and quantify high energy photon emitters. The technique chosen for the characterization of radioactive waste in the present work is the gamma-ray spectrometry with High purity Germanium (HPGe) detectors. From the energy identified in the experimental spectrum, three radioisotopes were identified in the cartridge filter: {sup 108m}Ag, {sup 110m}Ag, {sup 60}Co. For the estimated activity of the filter, the calibration in efficiency was made utilizing the MCNP4C code of the Monte Carlo method. Such method was chosen because there is no standard source available in the same geometry of the cartridge filter, therefore a simulation had to be developed in order to reach a calibration equation, necessary to estimate the activity of the radioactive waste. The results presented an activity value in the order of MBq for all radioisotopes. (authors)« less

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

The methodology is devised by coupling different codes. The study of weather conditions as part of the data of the site will determine the relative concentrations of radionuclides in the air using ARCON96. The activity in the air is characterized depending on the source and release sequence specified in NUREG-1465 by RADTRAD code, which provides results of the inner cloud source term contribution. Known activities, energy spectra are inferred using ORIGEN-S, which are used as input for the models of the outer cloud, filters and containment generated with MCNP5. The sum of the different contributions must meet the conditions of habitability specified by the CSN (Spanish Nuclear Regulatory Body) (TEDE <50 mSv and equivalent dose to the thyroid <500 mSv within 30 days following the accident doses) so that the dose is optimized by varying parameters such as CAGE location, flow filtering need for recirculation, thicknesses and compositions of the walls, etc. The results for the most penalizing area meet the established criteria, and therefore the CAGE building design based on the methodology presented is radiologically validated.

We are developing CHIPS-TPT physics library for exclusive simulation of neutron-nuclear reactions below 20 MeV. Exclusive modeling reproduces each separate scattering and thus requires conservation of energy, momentum and quantum numbers in each reaction. Inclusive modeling reproduces only selected values while averaging over the others and imposes no such constraints. Therefore the exclusive modeling allows to simulate additional quantities like secondary particle correlations and gamma-lines broadening and avoid artificial fluctuations. CHIPS-TPT is based on the formerly included in Geant4 CHIPS library, which follows the exclusive approach, and extends it to incident neutrons with the energy below 20 MeV. The NeutronHP model for neutrons below 20 MeV included in Geant4 follows the inclusive approach like the well known MCNP code. Unfortunately, the available data in this energy region is mostly presented in ENDF-6 format and semi-inclusive. Imposing additional constraints on secondary particles complicates modeling but also allows to detect inconsistencies in the input data and to avoid errors that may remain unnoticed in inclusive modeling.

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

During the decommissioning of nuclear facilities, a large number of cutting and demolition activities are performed, which results in a frequent change in the structure and produce many irregular objects. In order to assess dose rates during the cutting and demolition process, a flexible dose assessment method for arbitrary geometries and radiation sources was proposed based on virtual reality technology and Point-Kernel method. The initial geometry is designed with the three-dimensional computer-aided design tools. An approximate model is built automatically in the process of geometric modeling via three procedures namely: space division, rough modeling of the body and fine modeling of the surface, all in combination with collision detection of virtual reality technology. Then point kernels are generated by sampling within the approximate model, and when the material and radiometric attributes are inputted, dose rates can be calculated with the Point-Kernel method. To account for radiation scattering effects, buildup factors are calculated with the Geometric-Progression formula in the fitting function. The effectiveness and accuracy of the proposed method was verified by means of simulations using different geometries and the dose rate results were compared with that derived from CIDEC code, MCNP code and experimental measurements.

The nuclear accident at the Three Mile Island power plant in Pennsylvania has dealt yet another setback to the beleaguered nuclear power industry. The plant accident, combined with a recent renewal of interest in the 1974 Karen Silkwood case, increased pressure from regulatory agencies concerning plant safety and waste disposal, and the release of the anti-nuclear film ''''The China Syndrome'' has made the nuclear power industry a source of public criticism and consternation. The fact that officials at the Three Mile Island facility were unsure of the causes and amounts of the radiation leaks further adds to the predicament ofmore » the nuclear industry. The situation was compounded by the formation of a hydrogen gas bubble with the potential to cause a massive explosion at the plant. The incident has sparked protest rallies by anti-nuclear groups. Possible radiation exposure danger levels are assessed. (2 diagrams, 1 map, 9 photos)« less

The Nuclear Science References (NSR) database together with its associated Web interface, is the world's only comprehensive source of easily accessible low- and intermediate-energy nuclear physics bibliographic information for more than 210,000 articles since the beginning of nuclear science. The weekly-updated NSR database provides essential support for nuclear data evaluation, compilation and research activities. The principles of the database and Web application development and maintenance are described. Examples of nuclear structure, reaction and decay applications are specifically included. The complete NSR database is freely available at the websites of the National Nuclear Data Center (http://www.nndc.bnl.gov/nsr) and the International Atomic Energymore » Agency (http://www-nds.iaea.org/nsr)« less

The book begins with basic science and statistics relevant to nuclear medicine, and specific organ systems are addressed in separate chapters. A section of the text also covers imaging of groups of disease processes (eg, trauma, cancer). The authors present a comparison between nuclear medicine techniques and other diagnostic imaging studies. A table is given which comments on sensitivities and specificities of common nuclear medicine studies. The sensitivities and specificities are categorized as very high, high, moderate, and so forth.

This report summarizes the activities undertaken by EWI while under contract from the Department of Energy (DOE) Office of Nuclear Energy (NE) for the management and operation of the Nuclear Fabrication Consortium (NFC). The NFC was established by EWI to independently develop, evaluate, and deploy fabrication approaches and data that support the re-establishment of the U.S. nuclear industry: ensuring that the supply chain will be competitive on a global stage, enabling more cost-effective and reliable nuclear power in a carbon constrained environment. The NFC provided a forum for member original equipment manufactures (OEM), fabricators, manufacturers, and materials suppliers to effectivelymore » engage with each other and rebuild the capacity of this supply chain by : Identifying and removing impediments to the implementation of new construction and fabrication techniques and approaches for nuclear equipment, including system components and nuclear plants. Providing and facilitating detailed scientific-based studies on new approaches and technologies that will have positive impacts on the cost of building of nuclear plants. Analyzing and disseminating information about future nuclear fabrication technologies and how they could impact the North American and the International Nuclear Marketplace. Facilitating dialog and initiate alignment among fabricators, owners, trade associations, and government agencies. Supporting industry in helping to create a larger qualified nuclear supplier network. Acting as an unbiased technology resource to evaluate, develop, and demonstrate new manufacturing technologies. Creating welder and inspector training programs to help enable the necessary workforce for the upcoming construction work. Serving as a focal point for technology, policy, and politically interested parties to share ideas and concepts associated with fabrication across the nuclear industry. The report the objectives and summaries of the Nuclear Fabrication

A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

A method of absolute nuclear material assay of an unknown source comprising counting neutrons from the unknown source and providing an absolute nuclear material assay utilizing a model to optimally compare to the measured count distributions. In one embodiment, the step of providing an absolute nuclear material assay comprises utilizing a random sampling of analytically computed fission chain distributions to generate a continuous time-evolving sequence of event-counts by spreading the fission chain distribution in time.

ban on that flail away at each other in maneuvers designed to nuclear tests (sic), even for peaceful purposes, across build up influence . Whatever...uranium and develop reactors for nuclear can help Mr. Collor nudge (away from the nuclear submarines. program) his fractious military along by suspending...two and a half years to behaviour to qualify for a certificate. WASHINGTON permit the second six-year aid package for 1988-93, was POST has meanwhile

2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage casks and then convey title to the Secretary of Energy...far more economical options for reducing fossil fuel use .15 (For more on federal incentives and the economics of nuclear power, see CRS Report RL33442...uranium enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons

A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

"Nuclear Energy and the Environment" is a pocket folder of removable leaflets concerned with two major topics: Nuclear energy and Nuclear Techniques. Under Nuclear Energy, leaflets concerning the topics of "Radiation--A Fact of Life,""The Impact of a Fact: 1963 Test Ban Treaty,""Energy Needs and Nuclear Power,""Power Reactor Safety,""Transport,"…

This report presents the status at the end of 1989 and the outlook for commercial nuclear capacity and generation for all countries in the world with free market economies (FME). The report provides documentation of the US nuclear capacity and generation projections through 2030. The long-term projections of US nuclear capacity and generation are provided to the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) for use in estimating nuclear waste fund revenues and to aid in planning the disposal of nuclear waste. These projections also support the Energy Information Administration's annual report, Domestic Uranium Miningmore » and Milling Industry: Viability Assessment, and are provided to the Organization for Economic Cooperation and Development. The foreign nuclear capacity projections are used by the DOE uranium enrichment program in assessing potential markets for future enrichment contracts. The two major sections of this report discuss US and foreign commercial nuclear power. The US section (Chapters 2 and 3) deals with (1) the status of nuclear power as of the end of 1989; (2) projections of nuclear capacity and generation at 5-year intervals from 1990 through 2030; and (3) a discussion of institutional and technical issues that affect nuclear power. The nuclear capacity projections are discussed in terms of two projection periods: the intermediate term through 2010 and the long term through 2030. A No New Orders case is presented for each of the projection periods, as well as Lower Reference and Upper Reference cases. 5 figs., 30 tabs.« less

Entertains the thesis that social work has a stake in the technological-humanistic debate and should greet the recent and spectacular technological failures with protest and alarm. Discusses relationship of nuclear issue and social work, effects of nuclear issue on children, and Chernobyl. Advocates pacifism, activism, and a coherent conception of…

A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

This 1971 artist's concept shows the Nuclear Shuttle in both its lunar logistics configuraton and geosynchronous station configuration. As envisioned by Marshall Space Flight Center Program Development persornel, the Nuclear Shuttle would deliver payloads to lunar orbits or other destinations then return to Earth orbit for refueling and additional missions.

All the nuclear-weapon states are working to develop new nuclear-weapon systems and upgrade their existing ones. Although the US Congress has recently blocked further development of small nuclear weapons and earth-penetrating nuclear weapons, the United States is planning a range of new warheads under the Reliable Replacement Warhead programme, and renewing its nuclear weapons infrastructure. The United Kingdom is spending 1 billion pounds sterling on updating the Atomic Weapons Establishment at Aldermaston, and about 20 billion pounds sterling on replacing its Vanguard submarines and maintaining its Trident warhead stockpile. The US has withdrawn from the Anti-Ballistic Missile Treaty and plans to install missile defence systems in Poland and the Czech Republic; Russia threatens to upgrade its nuclear countermeasures. The nuclear-weapon states should comply with their obligations under Article VI of the Non-Proliferation Treaty, as summarised in the 13-point plan agreed at the 2000 NPT Review Conference, and they should negotiate a Nuclear Weapons Convention.

At the end of 1978, there were approximately 230 nuclear-fueled electric generating plants around the world; 72 of these were in the United States. Each plant requires an operations-and-maintenance workforce of 92 people, and attrition occurs at a rate of 8% per year. Requirements for a nuclear taskforce and job training, in view of current…

not be subject to International Atomic Energy Agency ( IAEA ) safeguards have the potential to produce 280...PNRA states that Pakistan follows IAEA physical protection standards. Proliferation A fundamental aspect of nuclear security is ensuring that...related to the design and fabrication of a nuclear explosive device,” according to the IAEA (Implementation of the NPT Safeguards Agreement in the

Presents an overview of the relation of nuclear power to human health and the environment, and discusses the advantages and disadvantages of nuclear power as an energy source urging technical educators to inculcate an awareness of the problems associated with the production of energy. Describes the fission reaction process, the hazards of…

The author recognizes a body of basic knowledge in nuclear power plant technoogy that can be taught in school programs, and lists the various courses, aiming to fill the anticipated need for nuclear-trained manpower--persons holding an associate degree in engineering technology. (Author/BP)

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

This paper reports on the development of the hypothesis of a nuclear dynamo for a controlled nuclear fusion reactor. This dynamo hypothesis suggests properties for a nuclear tornado that could annihilate nations if accidentally triggered by a single high yield to weight nuclear weapon detonation. The formerly classified reports on ignition of the atmosphere, the properties of a nuclear dynamo, methods to achieve a nuclear dynamo in the laboratory, and the analogy of a nuclear dynamo to a nuclear tornado are discussed. An unclassified international study of this question is urged.

A systematic macroscopic, leptodermous approach to nuclear statics and dynamics is described, based formally on the assumptions {h_bar} {yields} 0 and b/R << 1, where b is the surface diffuseness and R the nuclear radius. The resulting static model of shell-corrected nuclear binding energies and deformabilities is accurate to better than 1 part in a thousand and yields a firm determination of the principal properties of the nuclear fluid. As regards dynamics, the above approach suggests that nuclear shape evolutions will often be dominated by dissipation, but quantitative comparisons with experimental data are more difficult than in the case ofmore » statics. In its simplest liquid drop version the model exhibits interesting formal connections to the classic astronomical problem of rotating gravitating masses.« less

Described are the global effects of nuclear war. Discussed are radiation dosages, limited nuclear attacks, strategic arms reductions, and other results reported at the workshop on nuclear war issues in Moscow in March 1988. (CW)

On the 60th anniversary of the world's first nuclear power plant to produce electricity, Assistant Secretary for Nuclear Energy Peter Lyons discusses the Energy Department's and the Administration's commitment to promoting a nuclear renaissance in the United States.

Reports on a study that focused on the psychological impact of nuclear education curriculum on middle school students. Concluded that instruction about nuclear issues rarely increases students' fear or worry about nuclear war. (RT)

On the 60th anniversary of the world's first nuclear power plant to produce electricity, Assistant Secretary for Nuclear Energy Peter Lyons discusses the Energy Department's and the Administration's commitment to promoting a nuclear renaissance in the United States.

The topic of this survey article has seen a renaissance during the past couple of years. Here we present and extend the results for various phenomena which we have published from 2012-2014, with gratitude to our coauthors. The new phenomena include (a) the first reduced nuclear flux densities in vibrating diatomic molecules or ions which have been deduced from experimental pump-probe spectra; these "experimental" nuclear flux densities reveal several quantum effects including (b) the "quantum accordion", i.e., during the turn from bond stretch to bond compression, the diatomic system never stands still — instead, various parts of it with different bond lengths flow into opposite directions. (c) Wavepacket interferometry has been extended from nuclear densities to flux densities, again revealing new phenomena: For example, (d) a vibrating nuclear wave function with compact initial shape may split into two partial waves which run into opposite directions, thus causing interfering flux densities. (e) Tunneling in symmetric 1-dimensional double-well systems yields maximum values of the associated nuclear flux density just below the potential barrier; this is in marked contrast with negligible values of the nuclear density just below the barrier. (f) Nuclear flux densities of pseudorotating nuclei may induce huge magnetic fields. A common methodologic theme of all topics is the continuity equation which connects the time derivative of the nuclear density to the divergence of the flux density, subject to the proper boundary conditions. (g) Nearly identical nuclear densities with different boundary conditions may be related to entirely different flux densities, e.g., during tunneling in cyclic versus non-cyclic systems. The original continuity equation, density and flux density of all nuclei, or of all nuclear degrees of freedom, may be reduced to the corresponding quantities for just a single nucleus, or just a single degree of freedom.

There has been increased public interest in the potential effects of nuclear powerplant accidents since the Soviet reactor accident at Chernobyl. People have begun to look for more information about the amount of radioactivity that might be released into the environment as a result of such an accident. When this issue is discussed by people…

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and

Nuclear movement within a cell occurs in a variety of eukaryotic organisms including yeasts and filamentous fungi. Fungal molecular genetic studies identified the minus-end-directed microtubule motor cytoplasmic dynein as a critical protein for nuclear movement or orientation of the mitotic spindle contained in the nucleus. Studies in the budding yeast first indicated that dynein anchored at the cortex via its anchoring protein Num1 exerts pulling force on an astral microtubule to orient the anaphase spindle across the mother-daughter axis before nuclear division. Prior to anaphase, myosin V interacts with the plus end of an astral microtubule via Kar9-Bim1/EB1 and pulls the plus end along the actin cables to move the nucleus/spindle close to the bud neck. In addition, pushing or pulling forces generated from cortex-linked polymerization or depolymerization of microtubules drive nuclear movements in yeasts and possibly also in filamentous fungi. In filamentous fungi, multiple nuclei within a hyphal segment undergo dynein-dependent back-and-forth movements and their positioning is also influenced by cytoplasmic streaming toward the hyphal tip. In addition, nuclear movement occurs at various stages of fungal development and fungal infection of plant tissues. This review discusses our current understanding on the mechanisms of nuclear movement in fungal organisms, the importance of nuclear positioning and the regulatory strategies that ensure the proper positioning of nucleus/spindle. Published by Elsevier Ltd.

This work is critical study of literary interpretations of the nuclear holocaust. The author examines more than 250 stories and novels dealing with the theme of nuclear power and its devastating potential implications. Addressing such topics as the scientist and Armageddon, the role of religion, future evolution and mutation, and the postnuclear society, the author assesses the response of Bradbury, Lessing, Malamud, Shute, Huxley, Vonnegut, Heinlein, and others to the threat of nuclear apocalypse, with in-depth analyses of Alter Miller's A canticle for Leibowitz and Russell Hoban's Riddley Walker.

The fundamental capability of Nuclear Thermal Propulsion (NTP) is game changing for space exploration. A first generation Nuclear Cryogenic Propulsion Stage (NCPS) based on NTP could provide high thrust at a specific impulse above 900 s, roughly double that of state of the art chemical engines. Characteristics of fission and NTP indicate that useful first generation systems will provide a foundation for future systems with extremely high performance. The role of the NCPS in the development of advanced nuclear propulsion systems could be analogous to the role of the DC-3 in the development of advanced aviation. Progress made under the NCPS project could help enable both advanced NTP and advanced NEP.

Actin participates in several essential processes in the cell nucleus. Even though the presence of actin in the nucleus was proposed more than 30 years ago, nuclear processes that require actin have been only recently identified. Actin is part of chromatin remodeling complexes; it is associated with the transcription machineries; it becomes incorporated into newly synthesized ribonucleoproteins; and it influences long-range chromatin organization. As in the cytoplasm, nuclear actin works in conjunction with different types of actin-binding proteins that regulate actin function and bridge interactions between actin and other nuclear components. PMID:20452941

In the third annual report of the Carnegie Endowment for International Peace on the spread of nuclear weapons, Spector provides a critical survey of the status of nuclear proliferation throughout the world and examines the nuclear potential of nations in the Middle East, Asia, Africa, and Latin America. Drawing on both historical documents and up-to-date reports, the author addresses such specific topics as Israel's nuclear arsenal, nuclear terrorism and its global security implications, arms control and nuclear safeguards, international treaties, weapons buildup, and political radicalism and unrest in nuclear-threshold nations.

EPA’s RadNet system monitored the environmental radiation levels in the United States and parts of the Pacific following the Japanese Nuclear Incident. Learn about EPA’s response and view historical laboratory data and news releases.

Accurate cellular localization is crucial for the effective function of most signalling molecules and nuclear translocation is central to the function of transcription factors. The passage of large molecules between the cytoplasm and nucleus is restricted, and this restriction affords a mechanism to regulate transcription by controlling the access of transcription factors to the nucleus. In this Review, we focus on the signal transducer and activator of transcription (STAT) family of transcription factors. The regulation of the nuclear trafficking of STAT-family members is diverse. Some STAT proteins constitutively shuttle between the nucleus and cytoplasm, whereas others require tyrosine phosphorylation for nuclear localization. In either case, the regulation of nuclear trafficking can provide a target for therapeutic intervention.

RNA turnover is an integral part of cellular RNA homeostasis and gene expression regulation. Whereas the cytoplasmic control of protein-coding mRNA is often the focus of study, we discuss here the less appreciated role of nuclear RNA decay systems in controlling RNA polymerase II (RNAPII)-derived transcripts. Historically, nuclear RNA degradation was found to be essential for the functionalization of transcripts through their proper maturation. Later, it was discovered to also be an important caretaker of nuclear hygiene by removing aberrant and unwanted transcripts. Recent years have now seen a set of new protein complexes handling a variety of new substrates, revealing functions beyond RNA processing and the decay of non-functional transcripts. This includes an active contribution of nuclear RNA metabolism to the overall cellular control of RNA levels, with mechanistic implications during cellular transitions.

At Harvard President Derek Bok's request, six Harvard professors explain nuclear arms issues to help citizens understand all sides of the national security debates. The goal is to encourage public participation in policy formulation. The book emphasizes that escapism will not improve security; that idealistic plans to eliminate nuclear weapons are a form of escapism. Learning to live with nuclear weapons, they suggest, requires an understanding of the current nuclear predicament and the implications of alternative weapons and policy choices. After reviewing these matters, they emphasize that informed persons will continue to disagree, but that knowledge will improve understanding andmore » appreciation of their differences and improve the quality of policy debates. 54 references, 5 figures, 2 tables. (DCK)« less

As a result of the 1977 experience, it is shown that the U.S. no longer dominates the world nuclear market and must change its approach from coercion to persuasion. President Carter, implementing his campaign promises on nuclear nonproliferation, has used direct pressure, negotiated with nuclear suppliers, and asked for legislation to impose rigid criteria for the export of nuclear materials. Unilateral actions included the deferment of facilities for fuel reprocessing and breeder reactors, but were followed by efforts for international cooperation as the year progressed. While global non-proliferation policies reinforced with international technical cooperation are seen as admirable goals, themore » response to U.S. initiatives is not seen to be encouraging.« less

Over the past two decades, the biomechanical properties of cells have emerged as key players in a broad range of cellular functions, including migration, proliferation, and differentiation. Although much of the attention has focused on the cytoskeletal networks and the cell’s microenvironment, relatively little is known about the contribution of the cell nucleus. Here, we present an overview of the structural elements that determine the physical properties of the nucleus and discuss how changes in the expression of nuclear components or mutations in nuclear proteins can affect not only nuclear mechanics but also modulate cytoskeletal organization and diverse cellular functions. These findings illustrate that the nucleus is tightly integrated into the surrounding cellular structure. Consequently, changes in nuclear structure and composition are highly relevant to normal development and physiology and can contribute to many human diseases, such as muscular dystrophy, dilated cardiomyopathy, (premature) aging, and cancer. PMID:21756143

The densities of subgrade soil at various locations throughout the state were determined using the different model nuclear gages owned by the Department. In addition, some laboratory testing and sand cote testing were carried out. It was concluded th...

There is provided a direct nuclear pumped gas laser in which the lasing mechanism is collisional radiated recombination of ions. The gas laser active medium is a mixture of the gases, with one example being neon and nitrogen.

Nuclear power owes its origin to physicists. Fission was demonstrated by physicists and chemists and the first nuclear reactor project was led by physicists. However as nuclear power was harnessed to produce electricity the role of the engineer became stronger. Modern nuclear power reactors bring together the skills of physicists, chemists, chemical engineers, electrical engineers, mechanical engineers and civil engineers. The paper illustrates this by considering the Sizewell B project and the role played by physicists in this. This covers not only the roles in design and analysis but in problem solving during the commissioning of first of a kind plant. Looking forward to the challenges to provide sustainable and environmentally acceptable energy sources for the future illustrates the need for a continuing synergy between physics and engineering. This will be discussed in the context of the challenges posed by Generation IV reactors.

... into your blood and travels to your heart. Nuclear heart scans use single photon emission computed tomography (SPECT) or cardiac positron emission tomography (PET) to detect the energy from the tracer to make pictures of your ...

Future use of nuclear energy to produce electricity and desalted water is outlined. Possible desalting processes are analyzed to show economic feasibility and the place in planning in world's economic growth. (DS)

This presentation discusses: (1) Definitions of international nuclear security; (2) What degree of security do we have now; (3) Limitations of a nuclear security strategy focused on national lock-downs of fissile materials and weapons; (4) What do current trends say about the future; and (5) How can nuclear security be strengthened? Nuclear security can be strengthened by: (1) More accurate baseline inventories; (2) Better physical protection, control and accounting; (3) Effective personnel reliability programs; (4) Minimize weapons-usable materials and consolidate to fewer locations; (5) Consider local threat environment when siting facilities; (6) Implement pledges made in the NSS process; andmore » (7) More robust interdiction, emergency response and special operations capabilities. International cooperation is desirable, but not always possible.« less

Nuclear scintigraphy is a mainstay of diagnostic imaging and has preserved its relevance in the imaging of acute and chronic trauma. It is particularly useful in the evaluation of athletic injuries. Pitfalls of interpretation, false negatives and false positives exist as with many imaging modalities. Synthesis of physical exam findings, lameness evaluation and, when possible, diagnostic analgesia in combination with nuclear scintigraphy imaging findings, will allow for the most information to be applied to the patient's clinical problem. Published by Elsevier Inc.

Korea is developing nuclear weapons Yi and Bartholomew agreed that the two governments capability. should reestablish a bilateral science and...early this year. Taking advantage of such a mood of The government should develop state-of-the-art technol- detente. even our country is now actively...counter to such a mood of detente. and why can they not The report recommended that the government establish abandon nuclear weapons at a time when

A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

Using standard methods of nuclear optics, we investigate the recent suggestion by Kharzeev and Satz that antishadowing of photons in nuclear matter may be an effect of quantum interference. We show that the Kharzeev-Satz effect is controlled by the real part of the elastic scattering amplitude of the strongly interacting system coupled to the photon and traversing the nucleus. Phenomenological consequences of this observation are discussed.

missiles; cruise missiles; and gravity bombs . In contrast with the longer-range “strategic” nuclear weapons, these weapons had a lower profile in policy...States sought to deploy dual-capable aircraft and nuclear bombs at bases on the territories of NATO members in eastern Europe. Neither NATO, as an...ballistic missiles; cruise missiles; and gravity bombs . The United States deployed these weapons with its troops in the field, aboard aircraft, on

nuclear weapons program in 1941 and renamed the program the Manhattan Project in 1942.6 The mammoth efforts of the Manhattan Project resulted in the...killing or mortally wounding up to 130,000 Japanese.11 While the scientists of the Manhattan Project were awestruck at the first nuclear blast in New...remove great danger for us.”12 Klaus Fuchs and Theodore Hall, two scientists on the Manhattan Project , had been previously recruited to spy for the

The Lewis Research Center has a strong interest in nuclear rocket propulsion and provides active support of the graphite reactor program in such nonnuclear areas as cryogenics, two-phase flow, propellant heating, fluid systems, heat transfer, nozzle cooling, nozzle design, pumps, turbines, and startup and control problems. A parallel effort has also been expended to evaluate the engineering feasibility of a nuclear rocket reactor using tungsten-matrix fuel elements and water as the moderator. Both of these efforts have resulted in significant contributions to nuclear rocket technology. Many successful static firings of nuclear rockets have been made with graphite-core reactors. Sufficient information has also been accumulated to permit a reasonable Judgment as to the feasibility of the tungsten water-moderated reactor concept. We therefore consider that this technoIogy conference on the nuclear rocket work that has been sponsored by the Lewis Research Center is timely. The conference has been prepared by NASA personnel, but the information presented includes substantial contributions from both NASA and AEC contractors. The conference excludes from consideration the many possible mission requirements for nuclear rockets. Also excluded is the direct comparison of nuclear rocket types with each other or with other modes of propulsion. The graphite reactor support work presented on the first day of the conference was partly inspired through a close cooperative effort between the Cleveland extension of the Space Nuclear Propulsion Office (headed by Robert W. Schroeder) and the Lewis Research Center. Much of this effort was supervised by Mr. John C. Sanders, chairman for the first day of the conference, and by Mr. Hugh M. Henneberry. The tungsten water-moderated reactor concept was initiated at Lewis by Mr. Frank E. Rom and his coworkers. The supervision of the recent engineering studies has been shared by Mr. Samuel J. Kaufman, chairman for the second day of the

For 50 years, Los Alamos National Laboratory has been helping to keep nuclear materials secure. We do this by developing instruments and training inspectors that are deployed to other countries to make sure materials such as uranium are being used for peaceful purposes and not diverted for use in weapons. These measures are called “nuclear safeguards,” and they help make the world a safer place.

The investigations of the effects of nuclear transmutation in steels that are widely used in nuclear power and research reactors and in steels that are planned for the application in thermonuclear fusion plants, which are employed under the conditions of a prolonged action of neutron irradiation with different spectra, made it possible to study the effects of changes in the isotopic and chemical composition on the tendency of changes in the structural stability of these steels. For the computations of nuclear transmutation in steels, we used a program complex we have previously developed on the basis of algorithms for constructing branched block-type diagrams of nuclide transformations and for locally and globally optimizing these diagrams with the purpose of minimizing systematic errors in the calculation of nuclear transmutation. The dependences obtained were applied onto a Schaeffler diagram for steels used for structural elements of reactors. For the irradiation in fission reactors, we observed only a weak influence of the effects of nuclear transmutation in steels on their structural stability. On the contrary, in the case of irradiation with fusion neutrons, a strong influence of the effects of nuclear transmutation in steels on their structural stability has been noted.

During the Cold War, the United States and the Soviet Union competed in building weapons -- now it seems like America and Russia are competing to get rid of them the fastest. The lengthy process of formal arms control has been replaced by exchanges of unilateral force reductions and proposals for reciprocal reductions not necessarily codified by treaty. Should superpower nuclear strategies change along with force postures President Bush has yet to make a formal pronouncement on post-Cold War American nuclear strategy, and it is uncertain if the Soviet/Russian doctrine of reasonable sufficiency formulated in the Gorbachev era actually heraldsmore » a change in strategy. Some of the provisions in the most recent round of unilateral proposals put forth by Presidents Bush and Yeltsin in January 1992 are compatible with a change in strategy. Whether such a change has actually occurred remains to be seen. With the end of the Cold War and the breakup of the Soviet Union, the strategic environment has fundamentally changed, so it would seem logical to reexamine strategy as well. There are two main schools of nuclear strategic thought: a maximalist school, mutual assured destruction (MAD) which emphasizes counterforce superiority and nuclear war- fighting capability, and a MAD-plus school, which emphasizes survivability of an assured destruction capability along with the ability to deliver small, limited nuclear attacks in the event that conflict occurs. The MAD-plus strategy is based on an attempt to conventionalize nuclear weapons which is unrealistic.« less

In contrast to a history of support for our nation's involvement in wars, US churches are now in the process of edging toward a direct confrontation with the Reagan Administration over nuclear weapons and deterrence policy. This new attitude is attributed to two main factors: the growing awareness of the end of American strategic superiority and the concern over the Reagan Administration's attitude toward strategic nuclear issues. This newly widespread determination among the churches to challenge public policy on nuclear issues is addressed by examining the various churches and their efforts in the anti-nuclear movement. The Catholic Bishops' Draft Pastoralmore » Letter is discussed at length and is considered to be the most radical effort by any American church to define moral standards for the nuclear era. In contrast, the historical Protestant treatment differs considerably from the Catholic initiative; it started earlier but has been more fragmented and disparate. Only the American Jewish community as a whole has been, with a few notable exceptions, reluctant to become involved in the movement. The ultimate political question raised is how this religious involvement will influence American strategic nuclear doctrines. 27 references.« less

Potential paths to nuclear war and the available means of prevention of nuclear war are discussed. Presented is a detailed description of six nuclear war scenarios, and brief examples of types of potential deterrents to nuclear war (firebreaks) which are relevant for each. To be effective, the right combination of firebreaks must be used, the…

Suggests that science-related material on nuclear war be included in introductory courses. Lists nuclear war topics for physics, psychology, sociology, biology/ecology, chemistry, geography, geology/meteorology, mathematics, and medical science. Also lists 11 lectures on nuclear physics which include nuclear war topics. (JN)

Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. This paper examines the technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

character of conflict, arm control issues, conventional arms competition and U.S. forces; 3) Assess how new nuclear powers will behave and how their...neighbors 0and other nuclear powers will react; "--- 5) Identify the likely patterns and outcars of nuclear and other military interaction, including...Regional Nuclear Powers , 1990-2010 A small nuclear force (SNF) would comprise at a minimum from 5 to 10 deliverable and militarily serviceable fission

Nuclear Forensics is a growing field that is concerned with all stages of the process of creating and detonating a nuclear weapon. The main goal is to prevent nuclear attack by locating and securing nuclear material before it can be used in an aggressive manner. This stage of the process is mostly paperwork; laws, regulations, treaties, and declarations made by individual countries or by the UN Security Council. There is some preliminary leg work done in the form of field testing detection equipment and tracking down orphan materials; however, none of these have yielded any spectacular or useful results. Inmore » the event of a nuclear attack, the first step is to analyze the post detonation debris to aid in the identification of the responsible party. This aspect of the nuclear forensics process, while reactive in nature, is more scientific. A rock sample taken from the detonation site can be dissolved into liquid form and analyzed to determine its chemical composition. The chemical analysis of spent nuclear material can provide valuable information if properly processed and analyzed. In order to accurately evaluate the results, scientists require information on the natural occurring elements in the detonation zone. From this information, scientists can determine what percentage of the element originated in the bomb itself rather than the environment. To this end, element concentrations in soils from sixty-nine different cities are given, along with activity concentrations for uranium, thorium, potassium, and radium in various building materials. These data are used in the analysis program Python.« less

The first direct nuclear pumped laser using the He-2-(n,p) H-3 reaction is reported. Lasing took place on the 1.79 microns Ar I transition in a mixture of He-3-Ar at approximately 600 Torr total pressure. It was found that the electrically pulsed afterglow He-Ar laser had the same concentration profile as the nuclear pumped laser. As a result, nuclear lasing was also achieved in He-3-Xe (2.027 micron) and He-3-Kr (2.52 micron). Scaling of laser output with both thermal flux and total pressure as well as minority concentration has been completed. A peak output (He-3-Ar) of 3.7 watts has been achieved at a total pressure of 4 atm. Direct nuclear pumping of He-3-Ne has also been achieved. Nuclear pumping of a He-3-NF3 mixture was attempted, lasing in FI at approximately 7000 A, without success, although the potential lasing transitions appeared in spontaneous emission. Both NF3 and 238UF6 appear to quench spontaneous emission when they constitute more than 1% of the gas mixture.

The formation of complex nonuniform phases of nuclear matter, known as nuclear pasta, is studied with molecular dynamics (MD) simulations containing 51200 nucleons. A phenomenological nuclear interaction is used that reproduces the saturation binding energy and density of nuclear matter. Systems are prepared at an initial density of 0.10fm-3 and then the density is decreased by expanding the simulation volume at different rates to densities of 0.01fm-3 or less. An originally uniform system of nuclear matter is observed to form spherical bubbles (“swiss cheese”), hollow tubes, flat plates (“lasagna”), thin rods (“spaghetti”) and, finally, nearly spherical nuclei with decreasing density. We explicitly observe nucleation mechanisms, with decreasing density, for these different pasta phase transitions. Topological quantities known as Minkowski functionals are obtained to characterize the pasta shapes. Different pasta shapes are observed depending on the expansion rate. This indicates nonequilibrium effects. We use this to determine the best ways to obtain lower energy states of the pasta system from MD simulations and to place constraints on the equilibration time of the system.

This document has been prepared to incorporate all presentation aid material, together with some explanatory text, used during an oral briefing on the Nuclear Lunar Logistics System given at the George C. Marshall Space Flight Center, National Aeronautics and Space Administration, on 18 July 1963. The briefing and this document are intended to present the general status of the NERVA (Nuclear Engine for Rocket Vehicle Application) nuclear rocket development, the characteristics of certain operational NERVA-class engines, and appropriate technical and schedule information. Some of the information presented herein is preliminary in nature and will be subject to further verification, checking and analysis during the remainder of the study program. In addition, more detailed information will be prepared in many areas for inclusion in a final summary report. This work has been performed by REON, a division of Aerojet-General Corporation under Subcontract 74-10039 from the Lockheed Missiles and Space Company. The presentation and this document have been prepared in partial fulfillment of the provisions of the subcontract. From the inception of the NERVA program in July 1961, the stated emphasis has centered around the demonstration of the ability of a nuclear rocket to perform safely and reliably in the space environment, with the understanding that the assignment of a mission (or missions) would place undue emphasis on performance and operational flexibility. However, all were aware that the ultimate justification for the development program must lie in the application of the nuclear propulsion system to the national space objectives.

We have developed a comprehensive glossary of terms covering the broad fields of nuclear and related areas of science. The glossary has been constructed with two sections. A primary section consists of over 6,000 terms covering the fields of nuclear and high energy physics, nuclear chemistry, radiochemistry, health physics, astrophysics, materials science, analytical science, environmental science, nuclear medicine, nuclear engineering, nuclear instrumentation, nuclear weapons, and nuclear safeguards. Approximately 1,500 terms of specific focus on military and nuclear weapons testing define the second section. The glossary is currently larger than many published glossaries and dictionaries covering the entire field of physics. Glossary terms have been defined using an extensive collection of current and historical publications. Historical texts extend back into the 1800's, the early days of atomic physics. The glossary has been developed both as a software application and as a hard copy document.

Natural radioactive elements are part of our environment and radioactivity is a natural phenomenon. There are numerous beneficial applications of radioactive elements (radioisotopes) and radiation, starting from power generation to usages in medical, industrial and agriculture applications. But the risk of radiation exposure is always attached to operational workers, the public and the environment. Hence, this risk has to be assessed and controlled. The main goal of safety and security measures is to protect human life, health, and the environment. Currently, nuclear security considerations became essential along with nuclear safety as nuclear facilities are facing rapidly increase in cybersecurity risks. Therefore, prevention and adequate protection of nuclear facilities from cyberattacks is the major task. Historically, nuclear safety is well defined by IAEA guidelines while nuclear security is just gradually being addressed by some new guidance, especially the IAEA Nuclear Security Series (NSS), IEC 62645 and some national regulations. At the overall level, IAEA NSS 7 describes nuclear security as deterrence and detection of, and response to, theft, sabotage, unauthorized access, illegal transfer or other malicious acts involving nuclear, other radioactive substances and their associated facilities. Nuclear security should be included throughout nuclear facilities. Proper implementation of a nuclear security culture leads to staff vigilance and a high level of security posture. Nuclear security also depends on policy makers, regulators, managers, individual employees and members of public. Therefore, proper education and security awareness are essential in keeping nuclear facilities safe and secure.

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactormore » cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.« less

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

This symposium presented the latest techniques and approaches to the proper medical application of radionuclides in pediatrics. An expert faculty, comprised of specialists in the field of pediatric nuclear medicine, discussed the major indications as well as the advantages and potential hazards of nuclear medicine procedures compared to other diagnostic modalities. In recent years, newer radiopharmaceuticals labeled with technetium-99m and other short-lived radionuclides with relatively favorable radiation characteristics have permitted a variety of diagnostic studies that are very useful clinically and carry a substantially lower radiation burden then many comparable X-ray studies. This new battery of nuclear medicine procedures ismore » now widely available for diagnosis and management of pediatric patients. Many recent research studies in children have yielded data concerning the effacacy of these procedures, and current recommendations will be presented by those involved in conducting such studies. Individual papers are processed separately for the Energy Data Base.« less

In the early years of the United States space program, lightweight batteries, fuel cells, and solar modules provided electric power for space missions. As missions became more ambitious and complex, power needs increased and scientists investigated various options to meet these challenging power requirements. One of the options was nuclear energy. By the mid-1950s, research had begun in earnest on ways to use nuclear power in space. These efforts resulted in the first radioisotope thermoelectric generators (RTGs), which are nuclear power generators build specifically for space and special terrestrial uses. These RTGs convert the heat generated from the natural decay of their radioactive fuel into electricity. RTGs have powered many spacecraft used for exploring the outer planets of the solar system and orbiting the sun and Earth. They have also landed on Mars and the moon. They provide the power that enables us to see and learn about even the farthermost objects in our solar system.

The United States Supreme Court, with PG&E and Silkwood, and in the eight years since, has expanded the acceptable extent of state regulation of commercial nuclear power plants. In PG&E, the Court established the acceptability of state regulation that purports to be concerned with the non-radiological aspects of nuclear plant operations but that, as a practical matter, is concerned with their radiological hazards. In Silkwood, the Court established the acceptability of state regulation of radiological hazards when its impact on federal regulation of radiological hazards is indirect and incidental. Finally, in Goodyear and English, the Court confirmed and elaborated onmore » such state regulation. Subject to political demands either for additional involvement in commercial nuclear power plant regulation or from political interests opposed altogether to nuclear power, some states, in the 1980s, sought to expand even further the involvement of state and local governments in nuclear plant regulation. Indeed, some states sought and in some instances acquired, through innovative and extraordinary means, a degree of involvement in the regulation of radiological hazards that seriously erodes and undermines the role of the federal government in such regulation. In particular, the State of New York concluded with the Long Island Lighting Company (LILCO), in February 1989, an agreement for the purchase of New York of the Shoreham nuclear power plant on Long Island. A response to failed efforts by New York to prevent the issuance by the NRC of a license to LILCO to operate the plant, the agreement was concluded to allow New York to close the plant either altogether or to convert it to a fossil fuel facility. The opposition to the sale of Shoreham is discussed.« less

Widespread starvation resulting from changes in climate in the aftermath of a large-scale nuclear war could kill far more people than would the bombs themselves. That prediction was made in a recent study by the Scientific Committee on Problems of the Environment (SCOPE), an a rm of the International Council of Scientific Unions (ICSU). “Noncombatant and combatant countries alike” would risk mass starvation; SCOPE predicted that all told, 2.5 billion people could die as a result of crop failures and breakdowns in food distribution after a nuclear war.

The purpose of this book is to provide Americans with information and instructions that will significantly increase their chances of surviving a possible nuclear attack. It brings together field-tested instructions that, if followed by a large fraction of Americans during a crisis that preceded an attack, could save millions of lives. The author is convinced that the vulnerability of our country to nuclear threat or attack must be reduced and that the wide dissemination of the information contained in this book would help achieve that objective of our overall defense strategy.

The Nuclear Innovation Workshops were held at six locations across the United States on March 3-5, 2015. The data collected during these workshops has been analyzed and sorted to bring out consistent themes toward enhancing innovation in nuclear energy. These themes include development of a test bed and demonstration platform, improved regulatory processes, improved communications, and increased public-private partnerships. This report contains a discussion of the workshops and resulting themes. Actionable steps are suggested at the end of the report. This revision has a small amount of the data in Appendix C removed in order to avoid potential confusion.

The traveler was invited to serve as an external examiner for a doctoral thesis entitled Analysis of Myocardial Time-Activity Curves Related to Radiolabeled Free Fatty Acid Metabolism'' in the Cardiology Department at the Free University of Amsterdam, The Netherlands. The traveler also visited the Institute for Clinical and Experimental Nuclear Medicine in Bonn, West Germany, the Department of Nuclear Medicine in Aachen, West Germany, and the Cyclotron Research Center in Liege, Belgium. He led discussions, reviewed data, and coordinated further collaboration on the preclinical studies and clinical testing of radiopharmaceuticals being developed by the traveler's research group at the Oakmore » Ridge National Laboratory (ORNL).« less

It is an object of this invention to provide a nuclear imaging system having the versatility to do positron annihilation studies, rotating single or opposed camera gamma emission studies, and orthogonal gamma emission studies. It is a further object of this invention to provide an imaging system having the capability for orthogonal dual multipinhole tomography. It is another object of this invention to provide a nuclear imaging system in which all available energy data, as well as patient physiological data, are acquired simultaneously in list mode.

We analyzed in details the combined decay of the atomic-nuclear state, which consists of the excited 3/2{sup +} level of {sub 63}{sup 153}Eu and K hole, formed in the K capture by {sup 153}Gd. This decay proceeds in two stages. First, the nucleus transfers its energy to 2p electron, which flies into the continuum spectrum, and then returns into 1s hole, emitting γ quantum with the energy equal to the sum of energies of the nuclear and atomic transitions. We estimated the decay probability to be 2.2 × 10{sup −13}, that is much less than the recent experimental findings.

Nuclear Alert Forces. Four nuclear-armed states deploy nuclear warheads on alert, ready to be used on relatively short notice: United States, Russia, France and Britain. Combined, the four countries deploy an estimated 1,869 nuclear alert warheads. Russia and the United States deploy 1,749 alert warheads combined, or 94% of all alert warheads. Despite some debate about possible need to increase readiness of nuclear forces (China, Pakistan), the five other nuclear-armed states (China, Pakistan, India, Israel and North Korea) are thought to store their warheads separate from launchers under normal circumstances. The overall number of alert warheads has remained relatively stable during the past five years.

Although the outer nuclear membrane is continuous with the endoplasmic reticulum, it is possible to isolate nuclei both intact and free from endoplasmic reticulum contaminants. The outer and the inner nuclear membranes can be purified free from cross-contamination. Evidence in support of autonomous regulation of nuclear calcium signaling relies upon the investigations with isolated nuclei. Mechanisms for generating calcium signaling in the nucleus have been identified. Two calcium transporting systems, an ATP-dependant nuclear Ca(2+)-ATPase and an IP4-mediated inositol 1,3,4,5-tetrakisphosphate receptor, are located on the outer nuclear membrane. Thus, ATP and IP4, depending on external free calcium concentrations, are responsible for filling the nuclear envelope calcium pool. The inositol 1,4,5-trisphosphate receptor is located on the inner nuclear membrane with its ligand binding domain facing toward the nucleoplasm. Likewise, the ryanodine receptor is located on the inner nuclear membrane and its ligand cADP-ribose is generated within the nucleus. A 120 kDa protein fragment of nuclear PLC-gamma1 is stimulated in vivo by epidermal growth factor nuclear signaling coincident with the time course of nuclear membrane epidermal growth factor receptor activation. Stimulated 120 kDa protein fragment interacts with PIKE, a nuclear GTPase, and together they form a complex with PI[3]kinase serving as a module for nuclear PI[3]K stimulation. Thus, the nucleus has its own IP(3) generating system.

Briefly describes 46 courses on nuclear war available from United Campuses to Prevent Nuclear War (UCAM). These courses are currently being or have been taught at colleges/universities, addressing effects of nuclear war, arms race history, new weapons, and past arms control efforts. Syllabi (with assignments/reading lists) are available from UCAM.…

Professors Ernest Sternglass (University of Pittsburgh) and Steven Bell (Berry College) have assembled cogent, conclusive evidence indicating that nuclear radiation is associated with impaired cognition. They suggest that Scholastic Aptitude Scores (SATs), which have declined steadily for 19 years, will begin to rise. Their prediction is based on…

This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests inmore » NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.« less

A nuclear reactor core is described. It contains fuel in the form of blocks or pellets that have a grooved, wrinkled, or corrugated surface to provide a greater radiating surface area. The surfaces of spaces in the core are correspondingly corrugated for maximum heat exchange area. (C.E.S.)

The 101 nuclear plants operating in the US today are far safer than they were 20-30 years ago. For example, there's been about a 100-fold reduction in the occurrence of "significant events" since the late 1970s. Although the youngest of currently operating US plants was designed in the 1970s, all have been significantly modified over the years. Key contributors to the safety gains are a vigilant culture, much improved equipment reliability, greatly improved training of operators and maintenance workers, worldwide sharing of experience, and the effective use of probabilistic risk assessment. Several manufacturers have submitted high quality new designs for large reactors to the U.S. Nuclear Regulatory Commission (NRC) for design approval, and several companies are vigorously working on designs for smaller, modular reactors. Although the Fukushima reactor accident in March 2011 in Japan has been an almost unmitigated disaster for the local population due to their being displaced from their homes and workplaces and also due to the land contamination, its "lessons learned" have been important for the broader nuclear industry, and will surely result in safer nuclear plants worldwide - indeed, have already done so, with more safety improvements to come.

Presents a brief survey of the political and structural violence which pervades contemporary life. Attempts to demonstrate the ancient dictum that violence breeds more violence. Draws distinctions between different types of political violence and explores the political nature of nuclear deterrence. (JDH)

Energy consumption in Japan has grown at a faster rate than in any other major industrial country. To maintain continued prosperity, the government has embarked on a crash program for nuclear power. Current progress and issues/reactions to the plan are discussed. (JN)

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

The Nuclear Systems Kilopower Project was initiated by NASAs Space Technology Mission Directorate Game Changing Development Program in fiscal year 2015 to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (Technology Readiness Level 5) for space science and human exploration power needs. The Nuclear Systems Kilopower Project consists of two elements. The primary element is the Kilopower Prototype Test, also called the Kilopower Reactor Using Stirling Technology(KRUSTY) Test. This element consists of the development and testing of a fission ground technology demonstrator of a 1 kWe fission power system. A 1 kWe system matches requirements for some robotic precursor exploration systems and future potential deep space science missions, and also allows a nuclear ground technology demonstration in existing nuclear test facilities at low cost. The second element, the Mars Kilopower Scalability Study, consists of the analysis and design of a scaled-up version of the 1 kWe reference concept to 10 kWe for Mars surface power projected requirements, and validation of the applicability of the KRUSTY experiment to key technology challenges for a 10 kWe system. If successful, these two elements will lead to initiation of planning for a technology demonstration of a 10 kWe fission power capability for Mars surface outpost power.

The 101 nuclear plants operating in the US today are far safer than they were 20-30 years ago. For example, there's been about a 100-fold reduction in the occurrence of 'significant events' since the late 1970s. Although the youngest of currently operating US plants was designed in the 1970s, all have been significantly modified over the years. Key contributors to the safety gains are a vigilant culture, much improved equipment reliability, greatly improved training of operators and maintenance workers, worldwide sharing of experience, and the effective use of probabilistic risk assessment. Several manufacturers have submitted high quality new designs formore » large reactors to the U.S. Nuclear Regulatory Commission (NRC) for design approval, and several companies are vigorously working on designs for smaller, modular reactors. Although the Fukushima reactor accident in March 2011 in Japan has been an almost unmitigated disaster for the local population due to their being displaced from their homes and workplaces and also due to the land contamination, its 'lessons learned' have been important for the broader nuclear industry, and will surely result in safer nuclear plants worldwide - indeed, have already done so, with more safety improvements to come.« less

This 1-day colloquium, attended by 23 participants representing societies, government agencies, colleges and universities, and other training programs, was conducted for the purpose of reporting on and discussing the curriculums developed at the University of Cincinnati for training nuclear medical technologists. Pilot programs at both the…

Research has shown that nuclear radioisotope power generators can supply compact, reliable, and efficient sources of energy for a broad range of space missions. These missions range from televising views of planetary surfaces to communicating scientific data to Earth. This publication presents many applications of the advancing technology and…

The nuclear light bulb engine is a closed cycle concept. The nuclear light bulb concept provides containment by keeping the nuclear fuel fluid mechanically suspended in a cylindrical geometry. Thermal heat passes through an internally cooled, fused-silica, transparent wall and heats hydrogen propellant. The seeded hydrogen propellant absorbs radiant energy and is expanded through a nozzle. Internal moderation was used in the configuration which resulted in a reduced critical density requirement. This result was supported by criticality experiments. A reference engine was designed that had seven cells and was sized to fit in what was then predicted to be the shuttle bay mass and volume limitations. There were studies done of nozzle throat cooling schemes to remove the radiant heat. Elements of the nuclear light bulb program included closed loop critical assembly tests done at Los Alamos with UF6 confined by argon buffer gas. It was shown that the fuel region could be seeded with constituents that would block UV radiation from the uranium plasma. A combination of calculations and experiments showed that internal moderation produced a critical mass reduction. Other aspects of the research are presented.

High efficiency removal of technetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

When people think of the government they think of the President, or Congress, or the Internal Revenue Service (IRS), but there are thousands of people in government-related jobs doing things most don’t really notice everyday. You can find them everywhere, from the space science folks at NASA, to the Federal Bureau of Investigations (FBI) watching out for the bad guys. There are Rangers, and Social Workers, Nurses and Agricultural Managers. They are people working to keep the many facets of the USA rolling. One very diverse bunch is The Department of Energy (DOE) , a group who is expanding themore » ways we make and save energy to power our cars, homes, and businesses. Tucked away under the DOE is the National Nuclear Security Administration, the NNSA is an agency that maintains the safety, security, and reliability of the U.S. nuclear weapons stockpile. It works to reduce global danger from weapons of mass destruction. It provides the U.S. Navy with safe nuclear propulsion, and it responds to nuclear and radiological emergencies in the United States and abroad, and it supports efforts in science and technology*. (* DOE/NNSA/KCP website info)« less

The Committee on the Present Danger, Inc., the Committee of Atomic Bomb Survivors in the United States, the World Friendship Center in Hiroshima, two authors, physics and education professors, an English and history teacher, and a high school student comment on nuclear education. (RM)

Extensive advances in the technology of detectors, data analysis systems, and tracers used have resulted in greatly expanded applications of radioisotopes to the assessment of cardiac function and disease. The development of nuclear cardiology has proceeded along four lines: (1) radionuclide angiography, (2) myocardial perfusion imaging, (3) intracoronary microsphere imaging, and (4) regional myocardial blood flow determination using inert gases.

This report provides an overview of the scenarios used in strategic futures workshops conducted at Sandia on September 21 and 29, 2016. The workshops, designed and facilitated by analysts in Center 100, used scenarios to enable thought leaders to think collectively about the changing aspects of global nuclear security and the potential implications for the US Government and Sandia National Laboratories.

This book contains 19 chapters. Some of the titles are: Pulmonary Nuclear Medicine; Radionuclide Venography as an Adjunct to V-P Imaging in the Assessment of Thromboembolic Disease; Assessment of Mucous Transport in the Respiratory Tract by Radioisotopic Techniques; Radiolabeled Blood Cells and Tracers in the Study of Acute Pulmonary Injury and ARDS; and Magnetic Resonance Imaging of the Lungs.

... The special camera and imaging techniques used in nuclear medicine include the gamma camera and single-photon emission-computed tomography (SPECT). The gamma camera, also called a scintillation camera, detects radioactive energy that is emitted from the patient's body and ...